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METHODS FOR ASSESSINGOCCUPATIONAL RADIATION

DOSES DUE TOINTAKES OF RADIONUCLIDES

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METHODS FOR ASSESSING OCCUPATIONAL RADIATION

DOSES DUE TO INTAKES OF RADIONUCLIDES

INTERNATIONAL ATOMIC ENERGY AGENCYVIENNA, 2004

SAFETY REPORTS SERIES No. 37

IAEA Library Cataloguing in Publication Data

Methods for assessing occupational radiation doses due to intakes ofradionuclides. — Vienna : International Atomic Energy Agency,2004.

p. ; 24 cm. — (Safety reports series, ISSN 1020–6450 ; no. 37)STI/PUB/1190ISBN 92–0–103904–2Includes bibliographical references.

1. Ionizing radiation — Measurement. 2. Radiation dosimetry.3. Radiation workers. 4. Industrial safety. 5. Radiation — Safetymeasures. 6. Radioisotopes. I. International Atomic Energy Agency.II. Series.

IAEAL 04–00361

FOREWORD

Radioactive material is used in many human activities and, wheneverunsealed radioactive sources are present, intakes of radionuclides by workerscan occur. These activities include the use of radioactive sources in medicine,scientific research, agriculture and industry, the operation of various facilitiesthat are part of the nuclear fuel cycle, and work involving exposure to enhancedlevels of naturally occurring radionuclides. Intakes can occur by a number ofroutes, and the monitoring of workers and the workplace in such situations isan integral part of any occupational radiation protection programme.

Guidance on monitoring programmes and methods for assessments ofintakes of radioactive material arising from occupational exposure is given in aSafety Guide, Assessment of Occupational Exposure Due to Intakes ofRadionuclides (Safety Standards Series No. RS-G-1.2), published in 1999. Thisguidance is in turn supplemented by a Safety Practice on Direct Methods forMeasuring Radionuclides in the Human Body (Safety Series No. 114),published in 1996, and a Safety Report on Indirect Methods for AssessingIntakes of Radionuclides Causing Occupational Exposure (Safety ReportsSeries No. 18), published in 2000, that give practical advice on the methods forindividual monitoring of intakes of radionuclides by workers.

This report contains practical advice on the interpretation of monitoringresults and the assessment of committed effective doses to workers, using thestandard models of the International Commission on Radiological Protection,adopted as a reference in the International Basic Safety Standards forProtection against Ionizing Radiation and for the Safety of Radiation Sources(Safety Series No. 115), published in 1996. With its publication the IAEA nowprovides a complete set of reference publications for use in Member States thathave facilities in which workers have the potential to incur intakes ofradionuclides. These publications are founded on internationally acceptedprinciples and recommended practices, taking account of the major changes inprotection standards and monitoring methods that have occurred over the pastdecade.

This report was drafted over the course of five Consultants Meetings heldbetween 1997 and 2001 and finalized through a consultancy in 2003. The IAEAis grateful to the experts who took part in the development and review of thispublication. The contributions of J. Lipsztein, D. Noßke, A. Phipps, J.W.Stather, R.E. Toohey and D. Whillans are especially acknowledged. The IAEAofficer responsible for the preparation of this report was M. Gustafsson of theDivision of Radiation, Transport and Waste Safety.

EDITORIAL NOTE

Although great care has been taken to maintain the accuracy of informationcontained in this publication, neither the IAEA nor its Member States assume anyresponsibility for consequences which may arise from its use.

The use of particular designations of countries or territories does not imply anyjudgement by the publisher, the IAEA, as to the legal status of such countries or territories,of their authorities and institutions or of the delimitation of their boundaries.

The mention of names of specific companies or products (whether or not indicatedas registered) does not imply any intention to infringe proprietary rights, nor should it beconstrued as an endorsement or recommendation on the part of the IAEA.

CONTENTS

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1.1. Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11.2. Objective . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21.3. Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21.4. Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

2. BIOKINETIC MODELS FOR INTERNAL DOSIMETRY . . . . . . 4

2.1. Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42.2. Routes of entry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

2.2.1. Inhalation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72.2.2. Ingestion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 112.2.3. Wounds and intact skin . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

2.3. Systemic activity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 132.4. Excretion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 142.5. Dose coefficients. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 152.6. Workplace specific assessments . . . . . . . . . . . . . . . . . . . . . . . . . . 15

3. INTERPRETATION OF DIRECT AND INDIRECT MEASUREMENTS . . . . . . . . . . . . . . . . . . . . . . . . 16

3.1. Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 163.2. Assertions about m(t) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173.3. Application to interpretation of bioassay data . . . . . . . . . . . . . . 18

3.3.1. Assigning the time of intake . . . . . . . . . . . . . . . . . . . . . . . 183.3.2. Defining the route of intake . . . . . . . . . . . . . . . . . . . . . . . 193.3.3. Intake estimate from a single bioassay result . . . . . . . . . 193.3.4. Intake estimate from multiple bioassay data . . . . . . . . . 20

3.3.4.1. Point estimates . . . . . . . . . . . . . . . . . . . . . . . . . . . 213.3.4.2. Unweighted least squares fit . . . . . . . . . . . . . . . . 213.3.4.3. Weighted least squares fit . . . . . . . . . . . . . . . . . . 223.3.4.4. Maximum likelihood method and chi-square . . 22

3.3.5. Bayesian statistical inference . . . . . . . . . . . . . . . . . . . . . . 233.3.6. Intake estimates for extended exposures . . . . . . . . . . . . 23

3.3.6.1. Exposures over a time period . . . . . . . . . . . . . . . 243.3.6.2. Chronic and intermittent exposures . . . . . . . . . . 24

3.3.7. Interferences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 243.3.8. Intake estimates from measurements of related

nuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 253.4. Other dose assessment methods . . . . . . . . . . . . . . . . . . . . . . . . . . 25

3.4.1. Interpretation of air monitoring data . . . . . . . . . . . . . . . 253.4.2. Direct methods of dose calculation

without estimating intakes . . . . . . . . . . . . . . . . . . . . . . . . 263.5. Computer codes for dose assessment. . . . . . . . . . . . . . . . . . . . . . 263.6. Guidance for the design of monitoring programmes . . . . . . . . . 27

4. UNCERTAINTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28

4.1. Measurement results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 284.2. Intake characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 294.3. Biokinetic and dosimetric models . . . . . . . . . . . . . . . . . . . . . . . . 30

4.3.1. Biokinetic models. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 304.3.2. Dosimetric models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

4.4. Individual variations in biokinetic and dosimetric parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

4.5. Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34

5. DOSE RECORD KEEPING AND REPORTING . . . . . . . . . . . . . . 34

5.1. General considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 345.2. Individual monitoring records. . . . . . . . . . . . . . . . . . . . . . . . . . . . 345.3. Reporting information to management . . . . . . . . . . . . . . . . . . . . 35

6. QUALITY ASSURANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36

6.1. General considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 366.2. Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36

APPENDIX I: BASIC DATA FOR INTERNAL DOSE ASSESSMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39

APPENDIX II: BIOKINETIC MODELS FOR SELECTED ELEMENTS AND RADIONUCLIDES . . . . . . . . . . . . 47

APPENDIX III: RETENTION AND EXCRETION FRACTIONSFOR INTAKES OF SELECTED RADIONUCLIDES . . . . . . . . . . . . . . . . . 72

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75

ANNEX I: DETERMINING INTAKE FROM SINGLE AND FROM MULTIPLE DATA MEASUREMENTS FOR DOSE ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . 79

ANNEX II: DETERMINING THE TIME OF INTAKEFOR DOSE ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . 83

ANNEX III: DETERMINING THE ROUTE OF INTAKEFOR DOSE ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . 86

ANNEX IV: ANALYSIS OF AN INTAKE OF MIXED ACTIVATION AND FISSION PRODUCTS FOR DOSE ASSESSMENT . . . . . . . . . . . . . . . . . . . . . . . . . 90

ANNEX V: DOSE ASSESSMENT FROM AN EXPOSURE OVER A PERIOD OF TIME . . . . . . . . . . . . . . . . . . . . . . . 94

ANNEX VI: DIRECT DOSE ASSESSMENT FOR INTAKES OF TRITIATED WATER . . . . . . . . . . . . . . . . . . . . . . . . . . 97

ANNEX VII: ANALYSIS OF A SINGLE INTAKE OF 238,239,240Pu AND 241Am FOR DOSE ASSESSMENT . . . . . . . . . . . . . . 101

ANNEX VIII: CHOOSING THE APPROPRIATE MONITORING PERIOD FOR DOSE ASSESSMENT . . . . . . . . . . . . . . . . 109

GLOSSARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111CONTRIBUTORS TO DRAFTING AND REVIEW . . . . . . . . . . . . . . . 115

1

1. INTRODUCTION

1.1. BACKGROUND

Occupational exposure from intakes of radionuclides can occur as a resultof various activities, including: work associated with the different stages of thenuclear fuel cycle; the use of radioactive sources in medicine, scientificresearch, agriculture and industry; and occupations that involve exposure toenhanced concentrations of naturally occurring radionuclides. The 1990Recommendations of the International Commission on RadiologicalProtection (ICRP) [1], and the statutory requirements of most national author-ities, provide that for workers who are expected to be occupationally exposedto radioactive material an assessment be made of doses that may result fromintakes of radionuclides. Monitoring procedures are necessary if exposurescould arise that are subject to regulatory control.

Guidance on the protection of workers exposed to intakes of radio-nuclides is provided in the International Basic Safety Standards for Protectionagainst Ionizing Radiation and for the Safety of Radiation Sources (BSS) [2].The BSS give values of dose coefficients (doses per unit intake) for workersthat are the same as those issued in ICRP Publication 68 [3] and in theEuropean Union Directive on Basic Safety Standards [4].

For routine operations in which doses are likely to be small, thegeneralized biokinetic and dosimetric models that have been recommended bythe ICRP and incorporated into the dosimetry parameters contained in theBSS [2] are usually sufficient to provide a basis for the estimation of intakesand the determination of doses. In the case of accidents, however, or whenoperations could result in doses approaching regulatory limits, there may be aneed for dose estimates that are more specific to the individuals and theexposure situation, both for dose record keeping and for the effectivemanagement of those exposed. Detailed information on the physicochemicalform of the radioactive material, the exposure conditions, the retention charac-teristics of the radionuclides in the body, the anatomical and physiologicalcharacteristics of the individual or individuals involved and whether anytreatment was given may be needed in order that a more accurate doseassessment can be made.

The Safety Guide on Occupational Radiation Protection [5] gives generaladvice on the exposure conditions that would necessitate monitoringprogrammes for assessing radiation doses arising from external radiation andintakes of radionuclides by the workforce. The Safety Guide on Assessment ofOccupational Exposure Due to Intakes of Radionuclides [6] provides guidance

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on monitoring programmes and methods for assessing intakes of radionuclidestaken into the body by a worker. The latter Safety Guide is supported by theSafety Practice on Direct Methods for Measuring Radionuclides in the HumanBody [7] and the Safety Report on Indirect Methods for Assessing Intakes ofRadionuclides Causing Occupational Exposure [8]. A Safety Report ondosimetry services for individual monitoring, which is under development, willgive additional advice.

1.2. OBJECTIVE

The purpose of this report is to provide persons charged with the respon-sibility for monitoring internal exposures of workers with comprehensiveguidance on the methods for assessing committed effective doses fromestimated intakes of radioactive material, thereby supporting the informationgiven in Refs [7, 8]. However, in presenting the level of technical detailnecessary for this purpose, this report will also be useful to those concernedwith the planning and management of occupational monitoring programmes.

1.3. SCOPE

The aim of dose assessment for internal exposures is to obtain frommonitoring data estimates of committed effective doses or committedequivalent doses to individual organs or tissues. Monitoring data consist ofmeasurement data on levels of radionuclides in the whole body or in organsand tissues, or on their rates of excretion, or on their levels in the workenvironment, that can be used as a basis for assessing intakes and for relevantdose calculations.

This report presents the main considerations for dose assessment in bothroutine situations and accidents. Internal doses are computed by theapplication of biokinetic and dosimetric models to the results of direct orindirect monitoring measurements. Direct methods cover the measurement ofradiations emitted from radionuclides present in the body, whereas indirectmethods comprise activity measurements in biological samples, such as excreta,from which intakes of radionuclides can be calculated. Workplace sampling canalso provide supporting data, but is not reliable for accurate dose assessments.Effective doses are usually first determined from an estimate of the intakes ofradionuclides and application of the effective dose per unit intake coefficientsgiven in the BSS [2]. Where these estimates indicate that doses may be high, thedetermination may need to calculate doses using a specific dose model, which is

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beyond the scope of this report. In some circumstances dose rates may beassessed from measurements of body activity and used to infer doses withoutrelying on biokinetic models. This report offers practical advice on the inter-pretation of bioassay measurements in terms of intakes of radionuclides andthe resulting radiation effective doses for a number of radionuclides, using thestandard models recommended by the ICRP and adopted by the BSS [2].

A good programme of workplace radiation protection includes measuresother than the monitoring of workers; for example, it will include radiologicalengineering to provide the optimum engineered radiation protection, given themagnitude of the dose and the resources available, and workplace monitoringadequate to provide early and sensitive indicators of unexpected releases ofradioactive material. Guidance on these issues is beyond the scope of thisreport, but further details can be found in other IAEA reports [5, 9].

This report does not cover dose assessment for the medical exposure ofpatients or the exposure of members of the public. Neither does this report givespecific advice on dose assessment for workers exposed to radon (222Rn) orthoron (220Rn). Guidance on the assessment of occupational exposures to theseworkers is given in a Safety Guide on Occupational Radiation Protection in theMining and Processing of Raw Materials [10] and a Safety Report on RadiationProtection against Radon in Workplaces Other Than Mines [11].

1.4. STRUCTURE

The primary biokinetic and dosimetric models for describing thebehaviour of radionuclides in the body are summarized in Section 2, as are theorigin and use of the dose coefficients relating committed effective dose tointake. Section 3 describes the derivation and use of intake retention (orexcretion) functions to assess doses from the results of bioassay measurements,and also includes alternative dose assessment methods. Section 4 describes theuncertainties inherent in internal dose assessments and the possible techniquesfor minimizing or otherwise controlling them. Recommendations for recordkeeping and dose reporting are considered in Section 5. Finally, guidance onquality assurance procedures is given in Section 6. References and definitionsof terms used in internal dosimetry are also included.

The appendixes and annexes provide additional information. Appendix Iprovides basic data useful for internal dose assessment, including tables ofradiation and tissue weighting factors [1], and Appendix II contains biokineticmodels for selected elements. Appendix III, which is complemented by acompact disc attached to the back cover of this publication, provides datasheets that show, for selected radionuclides, values for the retention and

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excretion functions after intake. The values have been evaluated for varioustimes after intake. Annexes I–VIII contain detailed examples of doseassessments based on a variety of measurement data and exposure conditions.

2. BIOKINETIC MODELS FOR INTERNAL DOSIMETRY

2.1. INTRODUCTION

Intakes of radionuclides can occur via a number of routes. In the case ofoccupational exposure the main route of intake is by inhalation; a fraction ofmaterial deposited in the respiratory system will, however, be transferred tothe throat and swallowed, giving the opportunity for absorption in the gastro-intestinal (GI) tract. Intakes by ingestion may occur, as may absorptionthrough the intact skin for some radionuclides. Damage to the skin by cuts orother wounds can also result in intakes of radionuclides (Fig. 1).

In the case of workers who are occupationally exposed, the ICRP hasdeveloped a suite of models for describing the behaviour of radionuclides thathave entered the body either by inhalation or ingestion. For other pathways ofexposure, intakes are only likely to occur as a result of accidents that cannot becompletely prevented by workplace controls or readily predicted. No interna-tionally accepted models have therefore been developed that relate to eitherskin contamination or entry through wounds, although the National Council onRadiation Protection and Measurements (NCRP) is developing a model that isnearing completion. Some information on this issue has been published[12, 13]. A special case is tritiated water (HTO), which is readily absorbedthrough the skin; this may be assumed to be the route of intake for anadditional amount of tritium equal to 50% of the inhaled material, althoughthis is not a well founded value [3].

For intakes by ingestion, the GI tract model used to calculate the dosecoefficients given in the BSS [2] is that described in Ref. [14]. It describesmovement through four regions of the GI tract with parameter values forassessing the radiation dose to walls of the stomach and gut and the fractionaluptake of elements into the blood, given as f1 values (f1, the fractional uptakefrom the GI tract, is the gut transfer factor). For intakes by inhalation, theICRP [15] has described a Human Respiratory Tract Model (HRTM), whichhas replaced the lung model adopted in Ref. [14]. The HRTM takes account ofrecent information on the physiology of the lungs and is intended to beapplicable to the interpretation of bioassay data as well as the calculation of

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(a)

LiverTransfercompartment

Respiratorytract

model

Gastro-intestinal

tractmodel

Subcutaneoustissue

Lymphnodes

Otherorgans

Kidney

Urinary bladder

Skin

Wound

Sweat

Urine Faeces

IngestionExhalationInhalation

Direct absorption

Extrinsic removal

Skin

(b)Respiratorytract model tract model

Gastrointestinal

Tissuecompartment

1

Tissuecompartment

2

Tissuecompartment

3

Tissuecompartment

i

Gastrointestinaltract model

Urinarybladder

Transfer compartment

InhalationIngestion

Faecal excretion

Excretion

Systemic faecal excretionUrinary excretion

a2a1 a3 ai

fu ff

FIG. 1. (a) Routes of intake, transfer and excretion; (b) general model used to representthe kinetics of radionuclides in body compartments (exceptions are noted in the metabolicdata for individual elements) [6].

6

dose coefficients. It has been used for the calculation of inhalation dose coeffi-cients given in the BSS [2].

The current biokinetic models of the ICRP for systemic activity [16–19]were used for calculating dose coefficients given in the BSS [2] for intakes byinhalation and ingestion. These ICRP models comprise: (a) generic compart-ment models based on the models in Ref. [14] for some radionuclides; and (b)physiologically based recycling models for other radionuclides, all of which areconsidered in table III in Ref. [6]. This table III indicates, for each radionuclidelisted, the ICRP publication that was used as a source for each biokineticmodel.

In Ref. [20] the ICRP recommended the use of tissue weighting factors,wT, to calculate the committed effective dose equivalent from committed doseequivalents to individual tissues. This provided a mechanism for equating dosesand risks from external radiations, which are relatively uniform for all bodytissues, with those from intakes of radionuclides, which can be very hetero-geneous. In Ref. [1] and in the BSS [2] the approach used to calculate thecommitted effective dose equivalent (now termed ‘committed effective dose’)has been maintained, although as a result of improved information on the lateeffects of radiation on the tissues of the body some changes have been made tothe values of wT , and a greater number of tissues now have specified values. Atable of these tissue weighting factors is provided in Table 2.

Recommendations on methods for assessing doses from intakes of radio-nuclides from monitoring data have been made by the ICRP in Ref. [21], whichsupersedes Ref. [22]. These recommendations are based on the biokineticmodels developed by the ICRP and used for the calculation of dose coeffi-cients. These biokinetic models (which are summarized in Appendix II) canalso be used for the assessment of doses from indirect and direct measure-ments, in routine monitoring programmes, task related monitoring or specialmonitoring, when doses are low [6]. In these cases standard assumptions aboutthe time and pattern of intake, the physicochemical form of the radionuclidesand the characteristics of the individual (e.g. body mass) have to be used.However, when doses are greater than a few mSv a year, information from theworkplace needs to be gathered, for example the activity median aerodynamicdiameter (AMAD), as does more detailed information on the absorptioncharacteristics. The evaluation of doses in accident situations calls for morespecific information. Individual specific data on the biokinetics of radio-nuclides may be obtained through special monitoring (i.e. by repeated directmeasurements of the whole body or specific sites and measurements ofsequential biological samples).

The models used for calculating the inhalation and ingestion dose coeffi-cients for workers given in the BSS [2] and the dose coefficients for direct

7

uptake to blood (injection) given in this report are described below. Somediscussion of their use in assessing doses based on bioassay data is given,together with some comments on their limitations in different exposuresituations.

In the workplace, area air monitoring programmes are conducted toverify the effectiveness of the radioactive material containment, to detectaccidental releases and to provide a basis for the implementation of theinternal dosimetry programme. The control of airborne activity concentrationsin the workplace can be accomplished through the use of appropriate limits, thederived air concentrations (DACs) [6]. Note, however, that area air monitoringresults are unlikely to be sufficiently reliable to demonstrate compliance withthe DAC on an individual basis. For this application personal air samplers arelikely to be necessary.

2.2. ROUTES OF ENTRY

2.2.1. Inhalation

As in the earlier lung model [14], deposition and clearance are treatedseparately in the HRTM [15]. The scope of the HRTM includes all members ofthe population, giving reference parameter values for workers as well asmembers of the public, including infants and children.

Whereas the lung model given in Ref. [14] gives only the average dose tothe lungs, and no doses were calculated for extrathoracic (ET) tissues, doses tospecific tissues of the respiratory tract are calculated using the HRTM, takingaccount of perceived differences in their radiosensitivity. In the HRTM, therespiratory tract is represented by five regions (Fig. 2). The ET airways aredivided into ET1, the anterior nasal passage, and ET2, which consists of theposterior nasal and oral passages, the pharynx and the larynx. The thoracicregions are bronchial (BB), bronchiolar (bb) and alveolar–interstitial (AI), thegas exchange region. Lymphatic tissue is associated with the ET and thoracicairways LNET and LNTH, respectively.

Deposition of inhaled particles is calculated for each region of therespiratory tract, with account taken of both inhalation and exhalation. This isdone as a function of particle size, breathing parameters and/or workload, andis considered to be independent of chemical form. Regional depositionfractions are given for aerosols having log normal particle size distributions.Default deposition parameters are given for a particle size range of 0.6 nmactivity median thermodynamic diameter (AMTD) to 100 µm AMAD. For thepurpose of calculation, the reference worker is taken to be a normal nose

8

ExtrathoracicOral part

Nasal part

Anterior nasal passage

Posterior nasal passage

Pharynx

ET1

ET2Larynx

Trachea

BB Thoracic

Main bronchi

Bronchi

Bronchioles

Bronchial

bb

Al

Bronchioles

Bronchiolar

Alveolar –interstitial

Terminal bronchioles

Respiratory bronchioles

Alveolar duct + alveoli

bb

Al

FIG. 2. Respiratory tract regions defined in the new ICRP model (reproduced fromRef. [15]). The ET airways are divided into ET1, the anterior nasal passage, and ET2,which consists of the posterior nasal and oral passages, the pharynx and the larynx. Thethoracic regions are bronchial (BB: trachea and main bronchi), bronchiolar (bb: bron-chioles) and alveolar–interstitial (AI: the gas exchange region). Lymphatic tissue isassociated with the ET and thoracic airways (LNET and LNTH, respectively).

9

breathing adult male at light work. Regional deposition fractions for inhaledaerosols in the reference worker for a 5 µm AMAD aerosol are given inRef. [21]. This AMAD is now considered to be the most appropriate defaultparticle size for radionuclides in the workplace [15]. Inhalation dose coeffi-cients in the BSS (Ref. [2], table II-III) are given for an AMAD of 5 µm as wellas for an AMAD of 1 µm.

Clearance from the respiratory tract in the HRTM is treated as twocompeting processes: particle transport (by mucociliary clearance or trans-location to lymph nodes) and absorption to blood.

Particle transport is treated as a function of deposition site in therespiratory tract, but is taken to be independent of particle size and material.This time dependent mechanical transport is modelled by considering mostregions as a number of compartments with different clearance half-lives; forexample, the AI region is divided into three compartments, which clear to bbwith biological half-lives of about 35, 700 and 7000 days (Fig. 2). Similarly, bband BB have fast and slow clearance compartments. Clearance from the AIregion also involves transfer to lymphatic tissue. For bb, BB and ET2 there arecompartments to represent material that is sequestered in tissue or transportedto lymphatic tissue.

Absorption to blood depends on the physicochemical form of the radio-nuclide deposited in the respiratory system, but is taken to be independent ofdeposition site, with the exception of ET1, for which no absorption is assumedand activity is lost only by extrinsic means, such as nose blowing. The modelallows for changes in dissolution and absorption to blood with time. The use ofmaterial specific dissolution rates is recommended, but default parametervalues for the absorption rate to blood are given for use when no specificinformation is available. These are absorption Types F (fast), M (moderate)and S (slow), corresponding broadly to the default classes D, W and Y, respec-tively, in Ref. [14]. For all three absorption types, most of the materialdeposited in regions other than ET1 that is not absorbed is cleared to the GItract by particle transport. Small amounts transferred to lymph nodes continueto be absorbed into body fluids at the same rate as in the respiratory tract.

The transfer to blood for the different absorption types, when expressedas approximate half-lives, and the corresponding amounts of materialdeposited in each region that reach body fluids, can be summarized as follows:

(a) Type F: There is rapid absorption of almost all material deposited in BB,bb and AI. Half of the material deposited in ET2 is cleared to the GI tractby particle transport and half is absorbed. All the absorbed material isabsorbed with a biological half-life of 10 min. Examples are all thecommonly occurring compounds of caesium and iodine.

10

(b) Type M: There is rapid absorption of about 10% of the deposit in BB andbb, and 5% of material deposited in ET2. About 70% of the deposit in AIeventually reaches body fluids by absorption. Of the absorbed material,10% is absorbed with a biological half-life of 10 min and 90% with abiological half-life of 140 days. Examples are compounds of radium andamericium.

(c) Type S: There is little absorption from ET, BB or bb, and about 10% ofthe deposit in AI eventually reaches body fluids by absorption. Of theabsorbed material, 0.1% is absorbed with a biological half-life of 10 minand 99.9% with a biological half-life of 7000 days. Examples are insolublecompounds of uranium and plutonium.

For radionuclides inhaled in particulate form, it is assumed for workersthat entry into the respiratory system and regional deposition are governedonly by the size distribution of the aerosol particles. The situation is differentfor gases and vapours, for which deposition in the respiratory tract is materialspecific. Almost all inhaled gas molecules contact airway surfaces, but usuallyreturn to the air unless they dissolve in, or react with, the surface lining. Thefraction of an inhaled gas or vapour that is deposited in each region thusdepends on its solubility and reactivity. Generally, however, the regionaldeposition of a gas or vapour cannot be predicted on a mechanistic basis, fromknowledge of its physical and chemical properties, but has to be obtained froman experimental study in vivo.

In the HRTM [15] gases and vapours are assigned to three default classes,on the basis of the initial pattern of deposition in the respiratory tract:

(a) Class SR-0, insoluble and non-reactive: negligible deposition in therespiratory tract (e.g. 41Ar, 85Kr and 133Xe).

(b) Class SR-1, soluble or reactive: deposition may occur throughout therespiratory tract (e.g. tritium gas, 14CO, 131I vapour and 195Hg vapour).

(c) Class SR-2, highly soluble or reactive: total deposition in the ET airways(ET2) (e.g. HTO).

Subsequent retention in the respiratory tract and absorption to bodyfluids is determined by the chemical properties of the specific gas or vapour.

The approach taken by the ICRP and followed in the BSS [2] for givingguidance on the deposition and clearance of gases and vapours is similar to thatadopted for the clearance of radionuclides inhaled in particulate form. Forthose elements for which inhalation of radionuclides in a gaseous or vapourform is potentially important, defaults are recommended for the SR class(which determines deposition), and the corresponding absorption type (which

11

determines clearance), and may be Type V (very rapid absorption, for whichinstantaneous absorption is assumed for the purposes of calculation) or Type F(fast absorption). Consideration is given only to the behaviour of gases andvapours at low mass concentrations.

The dose coefficients for inhalation of particulate aerosols given intable II-III and those for inhalation of gases and vapours given in table II-IX ofthe BSS [2] are reproduced in Table 3.

2.2.2. Ingestion

The GI tract model used to calculate the dose coefficients given in theBSS [2] is that given in Ref. [14], with some modifications introduced to takeinto account the dose to the colon from systemic activity excreted in faeces andsome differences in the values of the fractional uptake from the GI tract. It is afour compartment model, comprising the stomach, small intestine, upper largeintestine and lower large intestine (Fig. 3). The mean residence times in the GItract compartments are 1, 4, 13 and 24 h, respectively. Absorption of radio-nuclides is usually assumed to occur from the small intestine. The contents ofthe small intestine are alkaline, so that elements that hydrolyse readily, such asrare earths and the actinides, are poorly absorbed. For the estimation of theequivalent dose to the wall of the GI tract from radionuclides in the lumen,doses are calculated to the mucosal cell layer. Note that the ICRP is currentlyin the process of developing a new model for the human alimentary tract [23].

The values for f1 given in the BSS [2] are reproduced in Table 3 and inAppendix II. The dose coefficients for ingestion given in table II-III of the BSS[2] are reproduced in Table 3.

2.2.3. Wounds and intact skin

Wounds and absorption through intact skin are additional routes bywhich radionuclides can enter the body. While much of the material may beretained at the wound site, soluble material can be transferred to the blood andhence to other parts of the body. Insoluble material will be slowly translocatedto regional lymphatic tissue, where it will gradually dissolve and eventuallyenter the blood. A variable fraction of insoluble material can be retained at thewound site or in lymphatic tissue for the life of the individual.

If the materials deposited in a wound are soluble, then they maytranslocate to the blood with a time course that depends on their dissolutionrate in vivo. The distribution of this soluble component will, in most instances,be similar to that entering the blood from the lungs or GI tract. The biokineticmodels developed by the ICRP can be used for the calculation of the effective

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Ingestion

Excretion

λB

λST

λSI

λULI

λLLI

Stomach (ST)

Small intestine (SI)

Upper largeintestine (ULI)

Lower largeintestine (LLI)

Body fluids

FIG. 3. Mathematical model used to describe the kinetics of radionuclides in the GI tract(reproduced from Ref. [14]).

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dose arising from the soluble component once the systemic uptake has beendetermined. As a first approximation, data for direct uptake to blood(injection) can be used. Note that these evaluations are very crude approxima-tions and are only to be used with caution, as an internationally agreed recom-mendation on models for wounds is still lacking.

A number of materials, such as specific tritium labelled compounds,organic carbon compounds and compounds of iodine, can penetrate intact skin.In these cases, a fraction of the activity will enter the blood directly. Althoughin many cases this activity will behave similarly to that which enters bloodfollowing inhalation or ingestion, there is some evidence [24, 25] that tritiatedorganic compounds can behave differently after absorption through skin. Carehas therefore to be exercised when assessing intakes through skin using theresults for direct uptake given in this report. Dose coefficients for injection aregiven in Table 3.

2.3. SYSTEMIC ACTIVITY

After translocation from the GI tract or lungs into body fluids (i.e. thetransfer compartment), an element is assumed to be cleared to organs, tissuesor excreta in accordance with the appropriate ICRP model (Appendix IIhereto). In general, a half-life in body fluids of 0.25 day is assumed.

The term ‘uptake’ refers to the process of translocation of material intothe systemic circulation, and may also refer to the quantity of material that hasentered the systemic circulation; this quantity, divided by the total amount ofmaterial in the intake, is called the ‘fractional uptake’.

A number of the revised systemic models for adults retain the genericmodel structure adopted in Ref. [14], but in some cases with minor changes tothe distribution of radionuclides between body compartments and theretention functions. In addition, for a number of elements, the models havebeen extensively revised, in particular to take account of recycling of radio-nuclides between compartments. To allow for the known behaviour of radio-nuclides and to take account of the present knowledge of the physiology ofbone, models for plutonium and other actinides (curium, americium,neptunium and thorium) and for the alkaline earths (calcium, strontium,barium and radium) have been developed [16–18]. The alkaline earth modelhas also been applied, with some modifications, to lead and uranium [17, 18].The specific ICRP publications from which the biokinetic models were takenare provided in table III in Ref. [6]. For the radionuclides included in this publi-cation, a summary of the systemic biokinetics adopted is given in Appendix II.

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A number of radionuclides decay to isotopes of elements that arethemselves radioactive. Generally it is assumed that the biokinetic behaviour ofthe decay products is the same as that of the parent nuclide. However, in thebiokinetic models for tellurium, lead, radium, thorium and uranium, separatesystemic biokinetics have been applied to the parent and its decay products[19].

2.4. EXCRETION

In the biokinetic models for workers, specific information is given onexcretion routes in urine and faeces. These models were adopted in Ref. [21],which gives information on the interpretation of bioassay data for selectedradionuclides based on the most recent biokinetic models.

For assessing doses from systemic activity excreted in the faeces, themodel for the GI tract is used (Section 2.2.2), assuming secretion of radionu-clides into the upper large intestine. Note, however, that this model was notdeveloped with the interpretation of bioassay data in mind: results aretherefore to be treated with caution. For urinary excretion a model for theurinary bladder has been adapted for calculating dose to the bladder wall [17].To represent the kinetics of the bladder in terms of first order processes, therate of elimination from the bladder is set equal to twice the number of voidsper day, taken to be six. That is, the elimination rate from the bladder is takento be 12 per day (equivalent to a biological half-life of about 1.4 h). This modelis used to derive the predicted daily urine excretion data given in this report.

Information is given by the ICRP on levels of excretion in urine andfaeces in the specific systemic models for different elements. This informationcan be given either as a ratio of urinary/faecal excretion (U:F), as forruthenium or zirconium, or as specific rate constants for loss in urine throughthe kidneys and in faeces by direct secretion into the lumen of the GI tract (e.g.plutonium and americium). Specific information on the excretion routes isgiven in Appendix II. Faecal excretion will reflect the unabsorbed fraction ofactivity entering the body as well as removal via the GI tract of systemicactivity. After inhalation of a radionuclide there will also be a component of(non-systemic) activity cleared directly from the respiratory system to the GItract.

15

2.5. DOSE COEFFICIENTS

Dose coefficients (committed effective doses per unit intake (Sv/Bq)) aregiven in the BSS [2] for intakes of radionuclides by ingestion and inhalation; itis always important to remember that dose coefficients refer to unit intake, not,for example, to unit deposition of activity in the respiratory tract followinginhalation intake. These values cannot be used directly for assessing doses fromdirect injection into the blood or from transfer to the blood from wound sites orabsorption through the skin. Additional dose coefficients for direct uptake toblood (injection), calculated using the models for systemic activity described inSection 2.3 and Appendix II, are therefore given in Table 3. Note that all valuesof committed effective dose include doses arising from any decay productsformed within the body.

For many radionuclides, dose coefficients are given for various lungabsorption types and different values of f1, the fractional absorption from theGI tract. In the BSS [2] advice is given that the most appropriate value of thedose coefficient must be based on knowledge of the physicochemical character-istics of materials in the workplace. Guidance is then given on the defaultvalues of f1 and lung absorption types for various chemical forms of theelements, which determine the appropriate dose coefficient (see Ref. [2],tables II-IV and II-V).

2.6. WORKPLACE SPECIFIC ASSESSMENTS

This report uses standard biokinetic models with default parametervalues and is only intended for assessing doses that may be considered to besmall (less than a few per cent of the dose limit [6]). A minimum of informationon the intake is necessary in order to use this report, namely: the radionuclidesthat may have been incorporated (including equilibrium assumptions for thenatural series), the chemical form of the compounds, their presumed aerosolsize (at least 1 or 5 µm) and the likely time frame, pattern and route of intake.

For exposures that lead to effective dose estimates higher than about5 mSv (a typical investigation level [6]), it will often be desirable to useparameter values in the calculation of tissue and organ equivalent dose that aremore specific to the conditions of exposure and to the individual.

By using such workplace specific parameters, a more realistic doseassessment can be obtained. However, a full discussion of the data andprocedures needed for such a task is beyond the scope of this report; furtherguidance is provided by the ICRP [26]. Readers are cautioned that the doseassessment from exposures to different uranium compounds, plutonium and

16

other transuranic elements usually calls for specialized help and that it isespecially important to keep track of the most recent information on thespecific biokinetic parameters for these elements, which are published by theICRP or can be found in the scientific literature.

3. INTERPRETATION OF DIRECTAND INDIRECT MEASUREMENTS

3.1. INTRODUCTION

Direct and indirect measurements provide information about theamounts of radionuclides present in the body, in parts of the body, includingspecific body organs or tissues, in a biological sample or in a sample taken fromthe work environment. The first approach to interpretation of these data islikely to be an estimation of the intake of the radionuclide by the worker.Biokinetic models (see Section 2) that describe body and organ contents andactivity in excreta as a function of time following intake, and exposure modelsthat relate intake to workplace conditions, are used for this purpose. Once theintake is estimated, the committed effective dose is then computed from theproduct of the intake and the appropriate dose coefficient. Alternatively, insome cases measurements of activity in the body can be used to estimate doserates directly, but the calculation of committed doses still calls for theassumption of a biokinetic model if an insufficient number of measurements isavailable to determine retention functions.

Figure 4 summarizes the general approach. In order to compute theestimated intake I, the measured body content, body region content orexcretion rate, M, is divided by the fraction of a unit intake, m(t), retained inthe whole body or in body regions (for direct in vivo measurements) orexcreted in a time period, usually per day, from the body (for excreta measure-ments) at time t (usually in days) after intake. Thus:

I = M/m(t)

Appendix III provides tables of values of m(t) for selected radionuclidesin urine, faeces, the whole body and selected tissues for intakes by inhalation of1 and 5 µm AMAD particles, inhalation of certain gases and vapours, ingestionand injection directly into the blood (transfer compartment). These values arebased on the biokinetic models described in Appendix II.

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When only a single bioassay measurement is available, a point estimate ofthe intake is made. If multiple measurements are available, a best estimate ofintake may be obtained by applying a statistical method. When significantintakes may have occurred, more refined calculations based on individualspecific parameter values (special dosimetry) need to be made.

3.2. ASSERTIONS ABOUT m(t)

Values of m(t) have been calculated from the models described inSection 2 and are tabulated in Appendix III and on the accompanying CD. Forwhole body or organ retention they give the fractional retention in the regionconsidered at time t (given in days). Some definitions are needed. Whole bodycontent is the sum of all systemic material, the contents of the urinary bladder,the GI tract and all regions of the respiratory tract. The contents of the lungs istaken to be the sum of the BB, bb and AI regions, together with the thoraciclymph nodes. The contents of the skeleton is taken to be the contents of thebone compartment in the simple models and the sum of all compartments ofcortical and trabecular bone and the bone marrow in the more complex models,for example those for plutonium and uranium.

For excretion, in general m(t) is the fraction of the intake excreted duringthe sampling period of 24 h preceding time t, taking into account radioactive

Directmeasurements

Indirectmeasurements

Body/organcontent, M

Doserate

Committedeffective dose

Estimatedintake

Excretion rate, M

Air concentration

m(t )m(t )

DAC-h

e(g )j

FIG. 4. General scheme for the interpretation of the results of monitoring measurements(possible alternative approaches for calculation are indicated as dashed lines) [6].

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decay. An exception is HTO, where urinary excretion values m(t) are given interms of activity concentration in urine at time t per unit intake.

The m(t) values in Appendix III are given for time since intake in days, onan expanding scale (i.e. t = 1, 2, 3,…, 10, 20, 30…, etc.). To obtain a value for atime not listed, a logarithmic interpolation between adjacent values is needed.If the values are not changing rapidly, a linear interpolation may be sufficientlyaccurate.

3.3. APPLICATION TO INTERPRETATION OF BIOASSAY DATA

For the application of the m(t) tables to the interpretation of bioassaydata, knowledge of the time and the route of the intake is necessary.

3.3.1. Assigning the time of intake

A principal source of uncertainty in the interpretation of bioassay data isthe determination of the time of intake. It is likely, especially for routinemonitoring, that the time will not be known beforehand. Since m(t) maychange rapidly with time, a reasonable estimate of the time of intake is vital forthe proper interpretation of bioassay data. If an unusual occurrence triggeredspecial bioassay monitoring, then the time of that occurrence is usually taken tobe the time of intake.

The most common approach when the time of intake is not known is toassume that it occurred at the midpoint of the monitoring period [20].However, if a significant intake and effective dose is calculated using thisassumption, then a more realistic determination is needed. Sometimes a reviewof workplace monitoring data, such as airborne or surface contamination levels,can indicate a likely time for the intake to have occurred. Similarly, if otherworkers in the same workplace have exhibited positive routine bioassaysamples, a review of the data and monitoring schedules for the individualworkers may help to determine the time of intake for all. Of course, anindividual worker may be able to recall the incident that led to the intake.

In addition, if several bioassay results are available, perhaps includingdifferent types of measurement, a comparison of these results with the m(t)tables may help in narrowing the period of time in which the intake occurred.

If the time of intake cannot be determined, and if there is evidence that aprolonged intake has occurred, then the bioassay data can be analysed byassuming a chronic intake situation, where, for a monitoring interval of n days,1/n of the total intake is assumed to have occurred each day. This method isdiscussed in Section 3.3.6.

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3.3.2. Defining the route of intake

Although intakes by inhalation alone are the most frequent in theworkplace, intakes by ingestion and uptake through wounds and intact skincannot be excluded. Sometimes a worker touches his or her mouth withcontaminated hands and ingestion occurs. If the route of intake is not knownand several bioassay results are available, including different types of bioassaymeasurements, a comparison of these results with the m(t) tables may help indetermining the route of intake. Occasionally, simultaneous intakes byinhalation and ingestion can occur. In principle, results from the ingestion andinhalation tables can be combined to give predicted values of m(t), but care willbe needed.

If the radionuclide activity can be assessed by direct measurements, lungcounting can be used to differentiate between inhaled and ingested material.However, if this is not possible and the radionuclide is in an insoluble form,interpretation of activities excreted in faecal and urine samples in terms ofintake is quite problematic. This is because both the inhaled material depositedin the upper respiratory tract and the ingested material will clear throughfaeces during the first few days after intake. Consequently, it is important toinitiate excreta sampling as soon as possible after the intake, and to continuemonitoring for an extended period. Material in the faeces after the secondweek will be exclusively from the respiratory tract, thus later measurementscan be used, together with the appropriate values of m(t), to correct themeasurements of earlier faecal samples so that they reflect ingestion intakesonly. It may be helpful to note that in the monitoring of workers chronicallyexposed to long lived, insoluble radionuclides, activities in faeces after anabsence of 15 days from work will mostly reflect delayed clearance frominhaled material [6].

Note that insoluble particles that have been inhaled and that are larger(e.g. AMAD > 15–20 µm) will preferentially be cleared through the GI tract,and so may present the appearance of an intake by ingestion. In this case,analysis of swabs of the nostrils and mouth may help to characterize the intakeroute.

3.3.3. Intake estimate from a single bioassay result

Once the time and route of intake have been determined or assumed, theintake, I, is simply given by:

I = M/m(t)

20

where

M is the bioassay result;t is the time since the intake; m(t) is the value of the bioassay prediction (retention or excretion) taken from

the tables given in this report.

Care must be taken to ensure that the bioassay result, M, and m(t) arecomparable; for example, in the case of urinalysis the bioassay result must beexpressed as the total activity in a 24 h urine sample at the end of collection(not at analysis).

Urine and faecal samples collected over periods of less than 24 h arenormalized to an equivalent 24 h value. This is achieved by multiplying thesample bioassay result by the ratio of the reference 24 h excretion volume (ormass) to the volume (or mass) of the sample. The reference volumes, for malesand females, respectively, are for urine 1.6 L and 1.2 L, and for faeces 140 g and120 g [27]. For urine, an alternative method is to normalize to the amount ofcreatinine excreted per day: 1.7 g for males and 1.0 g for females [8]. If the 24 hsample is less than 500 mL for urine or less than 60 g for faeces, then it isdoubtful that it has been collected over a full 24 h period, and normalizationneeds to be considered.

3.3.4. Intake estimate from multiple bioassay data

Usually, the bioassay data for an intake estimate will consist of results fordifferent samples collected at different times, and even from differentmonitoring techniques, for example urine data and faecal data, and perhapsalso from direct measurements. If the initial result from a routine sampleindicates a potentially significant intake, special monitoring is started to charac-terize the intake more accurately.

Numerous statistical methods are available for taking into accountmultiple measurements [28]. Most commonly used are a mean of the pointestimates and the unweighted least squares fit (ULSF). In some casesweighting factors are applied to give a weighted least squares fit (WLSF), andthis can result in a better fit to the data. A more general approach is theminimized chi-square method. Further information is given in the followingsections. When choosing a statistical method for multiple measurements, thevariability of the data and the reliability of each measurement need to be takeninto account; measurements that are suspect or known to be inaccurate need tobe excluded from the analysis. In short, the quality of the data set will influencethe reliability of the result.

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Caution must be used when combining data from different monitoringmethods; for example, if the results from direct measurements are severalorders of magnitude greater than the results from excreta analyses, the datashould not be combined in the methods described below. Ideally all measuredbioassay quantities should support consistent estimates of intake and dose. Thismay require model parameter values to be varied from their default values, butthey should not be varied beyond realistic bounds. These bounds will dependon the conditions of exposure; for example, the AMAD may not be known, butavailable information might indicate that the AMAD must lie between 1 and5 µm. If the predicted intakes from different bioassay data sets differ signifi-cantly, then each of the data sets needs a critical examination; for example, thedata could be unreliable or the model could be inappropriate in some respect.

3.3.4.1. Point estimates

In this method, each bioassay measurement is treated separately and isdivided by the appropriate predicted value, m(t), to generate a point estimateof the intake. If the resulting point estimates fall within a narrow range, a meanmay be taken to be the best estimate of the intake (see Annex I). However, ifthe point estimates of the intake fluctuate over a rather broad range of values,this is frequently an indicator that the standard biokinetic models used toderive the values of m(t), or the assumed time of intake, or the assumed routeof intake, may not be appropriate for this case. When dealing with excretasamples it is important to remember that there is a natural fluctuation in thedata, due to physiological factors and the influence of diet.

3.3.4.2. Unweighted least squares fit

The method used with the ULSF method is to minimize the sum of thesquares of the deviations of the observed measurements from the modelpredictions [28]. If Mi represents an observed value at time ti, I representsthe intake at time zero, and m(ti) is the value taken from the tables for time ti,the best fit to the data (in the least squares sense) is obtained by minimizing thesum S, where:

S = Σi[Mi – Im(ti)]2

The minimum can be found by differentiating S with respect to the intake, I,and setting the derivative equal to zero. The resulting expression can beexpanded and rearranged to give:

22

I = Σi[Mim(ti)]/Σi[m(ti)]2

This represents an unweighted least squares estimate of the intake.

3.3.4.3. Weighted least squares fit

Knowledge of the uncertainty for individual measurements will allowweighting of the importance of each point, or type of bioassay data, and henceweighting will influence the final result [28, 29]. Weighting factors (wi) could bechosen on the basis of a subjective assessment of confidence in each point ordata set, or by using some information on the error associated with eachmeasurement. In practice, the errors are likely to be unknown, and thereforesome type of assumption is usually adopted [29, 30]. This method needs to beused with caution.

The weighting factor may be applied at the initial sum of squares andcarried through the calculation, that is minimizing the sum S, where:

S = Σi[Mi – Im(ti)]2wi

which yields:

I = Σi[Mim(ti)wi]/Σi[m(ti)]2wi

3.3.4.4. Maximum likelihood method and chi-square

A more general approach is the maximum likelihood method (of whichthe ULSF and WLSF are special cases), which yields the most probable fit tothe data. The quantity to be minimized with respect to the estimate of intake, I,is known as chi-square, χ2, where:

χ2 = Σi[Mi – Im(ti)]2/σ2i

and σi is the standard deviation of Mi.In order to minimize χ2, this expression is differentiated with respect to I

and set equal to zero. Rearranging for I gives:

Clearly this method calls for some knowledge of the standard deviation,σi, for each measurement. Ideally this would be known from the measurement

I M m t m ti

n

i i ii

n

i i= ÂÏÌÓ

¸˝˛

ÂÏÌÓ

¸˝˛= =1

2

1

2 2[ ( ) [ ( ) ]/ ] /s s

23

procedure and knowledge of biological variability, but in practice it is usuallynecessary to make some assumption about σi; for example, it could be assumedthat σi is constant for each measurement, linearly related to the measurementitself (i.e. a constant relative error) or related to the square root of themeasurement [30]. Assumptions regarding σi can have a significant effect onthe estimate of intake, I, and need to be made with care [30, 31].

3.3.5. Bayesian statistical inference

A relatively recent (circa 1990) development in dose assessment is that ofBayesian statistical inference. The essence of the Bayesian approach is that itpresumes that the quantities of interest (in this case radionuclide intake anddose) are drawn from a probability distribution, rather than being fixed at someprecise, if unknown, value. One first assumes a plausible distribution for thequantities, referred to as a prior distribution, and then uses measured data toproduce an improved probability distribution, the posterior distribution. This isdone according to Bayes’s rule:

P(Θ|y) ~ P(y|Θ)P(Θ)

where P(Θ) is the prior probability distribution of the quantity Θ and P(Θ|y) isthe posterior probability distribution of Θ given measured data y. P(y|Θ) iscalled the likelihood function; it reflects the relative likelihood of obtaining themeasured data, given a particular value of Θ. A detailed explanation ofBayesian methods is beyond the scope of this report; further information isprovided in Refs [32–34].

3.3.6. Intake estimates for extended exposures

One of the factors that influence the interpretation of bioassay results isthe temporal variation of the intakes of radioactive material. The pattern ofintake, although often poorly characterized, is an important factor in thecorrect interpretation of measurements and thus for dose assessment. Ingeneral, the amount of activity present in the body and the amount excreteddaily depend on the length of time that the individual has been exposed. Conse-quently, the correct interpretation of bioassay measurements necessitatesinformation on the complete exposure history of the worker to the particularradionuclide of interest. The bioassay result obtained, for example the amountpresent in the body, in body organs or in excreta, will reflect the superpositionof all the previous intakes, whether isolated or persistent.

24

3.3.6.1. Exposures over a time period

When exposure is known to extend over several days, perhaps as a resultof an undetected incident, bioassay results may be interpreted as containing anindependent contribution from each day’s intake. Since the translocation ofradionuclides among body organs and tissues (compartments) is assumed to beindependent of previous intakes, the amounts expected to be present in thebody, in body organs or in excreta can be obtained by summing the m(t) values,perhaps weighted to allow for different daily intakes, for single intakes given inthis report. The results are added for each time period after single intakes, orfor the duration of a persistent intake.

3.3.6.2. Chronic and intermittent exposures

In routine monitoring of workers, especially for long lived radionuclides,it is highly desirable to produce a scheme in which the realistic exposure of theworkers (e.g. in a weekly cycle) is considered. The schedule of work may differfor individual workers, and modifications have to be introduced as necessary.The use of an input function that represents a worker’s routine intake permitsthe interpretation of bioassay results according to the day of the week on whichsamples are taken. In this way the short term components associated with lungclearance will be better accounted for, since the early clearance components ofexcretion may introduce a significant difference before and after an inter-ruption in exposure, for example over the weekend. The interpretation of thisdata in most cases calls for appropriate software tools and is beyond the scopeof this report.

For long lived radionuclides, chronic exposures that persist for some time,for example in a mining environment, can eventually produce an equilibriumvalue of activity in the body. Equilibrium values for selected radionuclides havebeen provided by the ICRP in Ref. [21].

3.3.7. Interferences

It is important to remember some factors that can lead to an inaccuracy ina dose assessment; examples include the presence of 137Cs from global fallout,radionuclides from the uranium and thorium series occurring naturally in thediet or radiopharmaceuticals that may have been administered for diagnosticor therapeutic purposes. Contamination on skin and on watches and jewellerycan also interfere with whole body and lung measurements. It is thereforeimportant to consider the body content and interferences from such intakes;more details are given in Ref. [6].

25

In addition, if medical intervention to prevent uptake or enhanceexcretion is carried out, note that the biokinetic behaviour will be modified andthat the tables of m(t) given in Appendix III may not be appropriate.

3.3.8. Intake estimates from measurements of related nuclides

Some radionuclides cannot be measured directly, but their body contentcan be assessed by the measurement of a daughter nuclide; examples are the invivo measurement of: 228Ac for the assessment of 232Th in the body; 214Bi for226Ra; and 234Th for 238U. These assessments rely on assumptions about theactivity ratios or equilibrium of the radionuclides or on a well establisheddegree of non-equilibrium [35, 36]. Values for m(t) for 228Ac in the bodyfollowing 232Th intakes are given in this report (Appendix III). Other radio-nuclides may be assessed by a related nuclide that is likely to be present in theintake, as for example the measurement of 241Am for 239Pu.

3.4. OTHER DOSE ASSESSMENT METHODS

3.4.1. Interpretation of air monitoring data

Samples from area air monitoring provide an indication of the radio-nuclides and their relative concentrations in the work environment. Thesesamples can provide a useful basis for determining the need for individualbioassay monitoring. Air samples may provide information on the particle sizedistribution and on the chemical form of the aerosol, which is needed for thecorrect interpretation of bioassay results. For intakes of some radionuclidesthat do not emit penetrating radiations, and that result in only very low levels inexcreta, personal air samples may be used to estimate the intake, whencombined with reasonable assumptions about the exposure conditions.

Air monitoring data can be used together with values of DACs takenfrom Table 4 and with estimated exposure times to give an indication of thelikely importance of an exposure. The DAC is that concentration of a radionu-clide in the air of the workplace that would result in workers receiving anintake that would produce an effective dose of 20 mSv if they were continu-ously exposed to the airborne activity for one working year (taken to be 2000h), breathing at a rate of 1.2 m3/h [6]. Care must be taken to ensure that theappropriate value of the DAC is used, as it depends on the particle size andHRTM absorption type of the radionuclide. In addition, the air beingmonitored must be representative of that which the workers are actually

26

breathing [8]. Area air monitoring results need to be calibrated againstpersonal air samplers before they can be reliably used in this application.

3.4.2. Direct methods of dose calculation without estimating intakes

In standard dose calculations based on intakes, assumptions concerningthe uptake, distribution and retention of the activity for a specific form andradionuclide have been established, and on this basis dose coefficients havebeen calculated [3]. However, there are a few cases for which doses may becalculated directly from the monitoring data, including urinary excretion datafor HTO, whole body data for 137Cs and thyroid data for 131I.

In short, this approach employs a simple numerical integration methodfor calculating the number of nuclear transformations (U) in the body (or someregion of the body) from measurement data. The trapezoidal rule (Fig. VI–1) isusually used to integrate a series of activity measurements, Mi, at times ti asfollows:

U = CΣi[(Mi+1 + Mi)/2](ti+1 – ti)

where C is a constant that reconciles any differences between the units of Mand t. Other more sophisticated integration methods could be used, but theerrors associated with the measurement data could outweigh any increases inaccuracy due to the method. A simple estimate of the additional transforma-tions beyond the last measurement (Mn) can be derived by assuming a defaulthalf-time (T) to estimate continuing retention:

U = CMn/[ln (2)/T ] or

U = 1.4CMnT

Committed doses can be calculated knowing the energies and yields ofthe emitted radiation and the masses of the regions where energy is absorbed.This approach is usually only practicable where the distributions of activity andthe pattern of energy depositions are simple. An example of this method beingapplied to HTO is given in detail in Annex VI.

3.5. COMPUTER CODES FOR DOSE ASSESSMENT

As noted in Section 2, the models describing the uptake, distribution andretention of radionuclides taken into the body are complex. The model

27

describing deposition and clearance, or absorption into the circulation, ofinhaled radionuclides may contain over 30 compartments itself, and thebiokinetic models for some elements may add another dozen. In addition, thetransport codes relating activity in source tissues to energy deposited in targettissues (absorbed dose), and the combination of these tissue doses, appropri-ately weighted for radiation quality (wR) and tissue sensitivity (wT) todetermine the effective dose, are computationally complex.

There are codes developed by specialized radiation dosimetry centres,such as the Bundesamt für Strahlenschutz (Germany), National RadiologicalProtection Board (UK) and Oak Ridge National Laboratory (USA), that areused to calculate the dose coefficients most recently published by the ICRP [3]and contained in the BSS [2]. These codes were used to calculate the values ofm(t) given in this report.

For the purposes of the industrial user, however, who has to determinedefault standards for workplace protection, and to estimate doses for particularincidents, a number of codes for personal computers have been developed thatare based on the recommendations of the ICRP. These codes typically containdefault models for the most common radionuclides and allow alternative valuesto be employed for some parameters, such as particle size and breathing rate.The codes then typically calculate retention and excretion for a variety ofradionuclides for different absorption types and intake routes, as well asabsorbed, equivalent and effective doses. They may also calculate intake anddose from multiple bioassay measurements using the statistical techniquesoutlined in Section 3.3.

Computer codes must be used with caution, however. It is important thatsoftware users maintain the configuration control of their software and operatethe software periodically. The software also needs to be benchmarked against awell understood intake with well characterized bioassay data, to determine if itproduces reliable results, for example in agreement with the m(t) values givenin this report.

3.6. GUIDANCE FOR THE DESIGN OFMONITORING PROGRAMMES

The IAEA has provided guidance on the design of monitoringprogrammes for intakes of radionuclides by workers [6], including advice onthe method and frequency of monitoring. It is important that the method usedfor monitoring has adequate sensitivity to detect the activity levels of interest.

The frequency and method of monitoring necessary in a routinemonitoring programme depend upon the activity being handled, the retention

28

and excretion of the radionuclide, the sensitivity of the measurementtechniques available and the acceptable uncertainty in the estimate of intakeand dose [21]. The magnitude and possible fluctuations of exposure levels alsoneed to be taken into account [6].

At some time after an intake, either the retained amount of activity in thebody or the amount being excreted from the body might be below theminimum significant activity (MSA: see Glossary) of the direct or indirectbioassay method. The values of m(t) provided in Appendix III can be useful todesign bioassay monitoring programmes, since values of intakes that could bemissed depending on the method and on the frequency of bioassay samplingcan be determined.

4. UNCERTAINTIES

In the process of assessment of radiation doses from intakes of radio-nuclides there are many sources of uncertainty. These are primarily due to:

(a) The uncertainty in the measurement results themselves;(b) Lack of knowledge about the intake characteristics (time pattern and

physical and chemical form);(c) The assumed biokinetic and dosimetric models;(d) The individual variability of biokinetic and dosimetric parameters.

4.1. MEASUREMENT RESULTS

Uncertainties in measurement results are discussed in the IAEA safetypublications on direct [7] and indirect [8] methods for measuring radionuclideswithin the body. There are no standard procedures for indirect or directbioassay measurements, although some examples of bioassay methods aregiven in these publications. The choice of the procedure, detector or facility willdepend on the specific needs, such as the nuclides of interest, minimumdetectable activities, budget, etc. All procedures used to quantify the activitiesof a radionuclide are sources of random and systematic errors. Uncertainties inmeasurements are mainly due to counting errors, the validity of the calibrationprocedures, possible contamination of the source or the measurement systemand random fluctuations in the background.

29

Examples of sources of uncertainty for in vitro measurements include: thequantification of the sample volume or weight; errors in dilution and pipetting;evaporation of the solution in storage; stability and activity of standards usedfor calibration; similarity of chemical yield between the tracer and radio-element of interest; blank corrections; background contributions and fluctua-tions; electronic stability; spectroscopy resolution and peak overlap;contamination of the sample and impurities; source positioning for counting;density and shape variation from the calibration model; assumptions abouthomogeneity in calibration; and statistical counting errors [37].

For in vivo monitoring, common sources of uncertainty include: countinggeometry errors; positioning of the individual in relation to the detector;movement of the individual during counting; chest wall thickness determi-nation; differences between the phantom and individual or organ beingmeasured, including geometric characteristics, density, distribution of theradionuclide within the body and organ and linear attenuation coefficient;interference from radioactive material deposits in adjacent body regions;spectroscopy resolution and peak overlap; electronic stability; interferingbackground activity and interference from other radionuclides; backgroundstability; activity of the standard radionuclide used for calibration; surfaceexternal contamination of the person; interference from natural radioactiveelements present in the body; counting statistics during calibration and duringin vivo counting; and calibration source uncertainties [7, 37].

For partial body measurements, it is difficult to express the result in termsof organ activities. For the determination of lung activity by measurement overthe chest, for example, not only individual calibration problems (such as thethickness of the individual’s chest wall) must be considered, but also radiationfrom various other body regions, not only from the lungs, may be detected. Soadditionally some assumptions must be made about the biokinetic behaviour ofthe radionuclide. An example for the case of 241Am is given in Ref. [7].

4.2. INTAKE CHARACTERISTICS

For the interpretation of direct and indirect measurements in terms of theintake and resulting effective dose, data on the time pattern and route ofintake, on the chemical and physical form of the radionuclides and on previousintakes are needed. In many cases this information is not available.

The time pattern is a main source of uncertainty in the interpretation ofbioassay data. Assumptions about the time of intake, and of whether the intakewas acute, lasted for a short period of time or extended for a long time, are amajor point in the reliability of the interpretation of the bioassay data. For

30

example, in some cases the retention and excretion functions diminish byorders of magnitude within a few days; the choice of the time pattern of intakecan therefore influence the assessed dose within the same range (see tables ofm(t) in Appendix III). Inhalation is the main route of intake. Characterizationof the intake in terms of aerosol size and absorption type is needed for theapplication of the m(t) values. The aerosol size will influence deposition in theHRTM and the transfer of unabsorbed particles to the GI tract. In some workenvironments more than one particle size is detected. As a minimum, m(t)values are given for particle sizes of 1 and 5 µm. The rate of absorption of aradionuclide to blood is very important for interpreting bioassay data, and is acritical parameter in interpreting urine excretion data. The differences betweenthe true absorption rates and the default parameter values that have beenassigned to the compound being inhaled are a source of errors that can be verylarge, especially when deriving intakes from urinary excretion bioassay data.

Further uncertainty is added when the activity of a radionuclide in thebody cannot be measured directly but is derived from progeny radionuclides(see Section 3.3.8).

Another source of uncertainty is the assignment of a route of intake. Inmany circumstances there is a mixture of ingestion and inhalation, and inter-pretation of results based on wrong assumptions about the pathway ofexposure may lead to large errors in interpreting in vitro bioassay results.

Contributions from intakes from natural sources, especially in the diet,may also contribute to the uncertainty of a bioassay result.

4.3. BIOKINETIC AND DOSIMETRIC MODELS

4.3.1. Biokinetic models

The biokinetic models used in this report comprise the most recentmodels published by the ICRP. Important advances in the models have beenmade in recent years, and have incorporated increased physiological realism.Developments in this area are ongoing: for example the modifications in themodel for the human alimentary tract [23].

The reliability of the biokinetic models is associated with uncertainties inthe sources and the quality and completeness of data used in the derivation ofthe models [38]. These uncertainties include the stochastic variability and thelack of knowledge about a single true value or a true but unknown distributionof values. The reader is referred to Refs [38, 39] for a detailed discussion on thereliability of biokinetic models. Biokinetic parameter values are derived fromdirect observations of the time distribution and excretion of the element in

31

humans and from analogies of the behaviour of the element in other species, ofthe biokinetics of chemically similar elements in humans and animals, and ofthe in vitro behaviour of the element of interest [39]. This information issometimes supplemented by considerations of mass balance and physiologicaldata [38].

While there are some elements for which extensive human data areavailable to develop reliable models, there are also many elements for whichthe confidence in the model is relatively low; for example, to determine the GIabsorption factor (f1) for antimony, there are only animal data available that liewithin a range from less than 0.01 to 0.2, depending on the chemical form of theelement, and a value of 0.1 was chosen by the ICRP for ingested antimony [18,40], but clearly there is a large degree of uncertainty attached to this parameter.In addition, inhalation is the most important route of intake of radionuclides inthe human body. In spite of the major advance in the model structure for therespiratory tract, there are areas of uncertainty, which lead to differences in thedefault parameter values adopted by the ICRP [15] and by the NCRP [41]. Thedefault parameter values for the respiratory tract model often do not representa particular compound accurately. The ICRP recommends [15] that materialspecific rates of absorption be used for compounds for which reliable experi-mental data exist. The use of absorption material specific rates for importantcompounds is discussed and illustrated by examples in Ref. [25], while onlydefault absorption parameter values were used in the derivation of the m(t)values for this report. Differences in deposition, retention and dosimetry of theICRP and NCRP respiratory tract models are discussed in Ref. [42].

Even in cases in which extensive human data are available, there still maybe much uncertainty in the estimated effective dose if the assumed biokineticmodel does not consider all relevant components of the actual biokineticbehaviour of the radionuclide; for example, it is accepted that the whole bodyretention of caesium can be well described by the sum of two exponentialfunctions with biological half-lives of about two and 110 days, respectively, ashas been done by the ICRP [14, 16]. However, the data supporting this modelwere only collected for a few days to a few months after intake. Longer termdata obtained following the Goiânia incident [43], however, indicate that thereis an additional, small, long term component (about 0.1% of initial systemicactivity) with a biological half-life of about 500 days. This third component hasno relevant influence on the effective dose per unit intake, but for theinterpretation of bioassay measurements at long times after an intake it mayinfluence the estimated intake activity, and therefore the resulting effectivedose, by an order of magnitude.

Although inhalation and ingestion are the most important pathways ofinternal contamination, absorption through wounds and intact skin may also

32

occur. While the use of m(t) values, derived for direct uptake into blood(injection), can be used to assess intakes of soluble materials from wounds,note that this could present a significant source of uncertainty.

The models used to derive the m(t) values in this report have beendesigned to be applied to a typical or reference individual, who might beexposed to low level internal contamination in his or her duties at work. Theyare meant to ensure the adequacy of radiological controls. In cases in whichthere is known to be a high degree of uncertainty associated with a model, it isadvisable that the user verify that the model is being used to interpret measure-ments that are roughly in the same range as the data on which the model wasbased. In such a range, a model put forward in this report may be accepted asreliable.

4.3.2. Dosimetric models

Beyond the uncertainties in the standard biokinetic models discussedabove, there are also uncertainties in the models that describe the energydeposition in the target regions. These include uncertainties due to the use ofsimplifying assumptions about organ masses, sizes and shapes, and due to thegeometrical relationships between internal organs that are implicit in the use ofcomputer phantoms. There are also limitations to the computationalprocedures for the calculation of specific absorbed fractions for penetratingradiations, and in the simplified assumptions about absorbed fractions in thebone and GI tract for non-penetrating radiations.

4.4. INDIVIDUAL VARIATIONS IN BIOKINETIC ANDDOSIMETRIC PARAMETERS

The biokinetic and dosimetric models are designed for a referenceindividual, that is for an individual representing average values for the groupconsidered, such as the Reference Man [27]. There are, however, considerabledifferences among individuals of such a group. Variations in anatomical andphysiological factors influence the distribution and excretion of radionuclidesin the body. These are, for example, differences in genetic constitution, age, sex,breathing patterns, lung, renal, liver, GI and cardiovascular functions,pregnancy and lactation. Environmental factors such as exercise, disease, stress,infection, smoking, alcohol intake, dietary factors, barometric pressure andexposure to sunlight may interact with biological factors, producing sizablevariations among individuals [39]. Examples of biological and environmentalinfluences are given below.

33

Different individuals ingesting lead under similar conditions have shownfractional GI absorption values of between 0.01 and 0.16 [44]. It is thereforeinevitable that individual parameter values may be significantly different fromthe average values used in the standard biokinetic models.

Iodine uptake by the thyroid is largely dependent on the stable iodinealready in the thyroid, which is influenced by the amount of stable iodine in theindividual’s diet. Therefore, in countries with low levels of iodine in typicalfoodstuffs, a higher iodine uptake is observed (45–50% instead of the ICRPvalue of 30%) [45]. Since this higher uptake is frequently correlated with ahigher thyroid mass, this does not influence the thyroid dose and the effectivedose per unit intake significantly. However, this higher uptake greatlyinfluences the estimated intake and effective dose based on bioassay data.

The nominal daily urinary excretion from Reference Man is 1.6 L, butsince this depends strongly on physiological and environmental conditionsthere can be very large variations in the excreted activity from one day toanother for the same individual, which cannot easily be interpreted by abiokinetic model. For the interpretation of excretion measurements throughthe m(t) tables of this report, a 24 h sample is preferred (with the exception ofHTO); unfortunately, this cannot be assured in all cases, and the normalizationof data to 24 h excretion will be another source of uncertainty.

Faecal samples from individual voidings vary widely in mass, compositionand transit time through the GI tract. In addition, they are very difficult tointerpret, since they contain materials cleared from the lungs, systemic materialexcreted into the GI tract and material passing unabsorbed through the GItract following ingestion. Each of these variables are sources of uncertaintywhen using the values of m(t) for faecal excretion to derive the intake. Formany bioassay monitoring programmes, samples need to be collected over athree day period to estimate the daily excretion rate. Often the complete 24 hexcretion is not available or the worker cannot provide samples for a period ofseveral days, and normalization to the Reference Man excretion rate isnecessary, which introduces another source of error.

An individual specific analysis is only necessary for the few situations inwhich the worker’s dose approaches the dose limit. Even when specific analysisis conducted, the day to day variation, the behaviour of the individual inrelation to environmental factors and the limitations of measurements willintroduce uncertainties that are not easily quantified.

34

4.5. SUMMARY

The overall uncertainty in assessed dose is a combination of the uncer-tainties noted in the previous sections. A reliable estimate of this overalluncertainty is, however, difficult to achieve. The ICRP recommends making adose assessment on the standard basis and adopting the results as nominalvalues of intake and dose [21]. If the dose assessed in this way is significant,then the uncertainties need to be considered in more detail.

5. DOSE RECORD KEEPING AND REPORTING

5.1. GENERAL CONSIDERATIONS

Dose record keeping is the creation and maintenance of individual doserecords for radiation workers. It is an essential part of the process ofmonitoring exposures of individuals to radiation and supports the overallobjectives of the radiation protection programme. General guidance is given inRef. [5]. Further information that relates to doses from intakes of radionuclidesis given below.

5.2. INDIVIDUAL MONITORING RECORDS

Typical records generated in an internal exposure monitoring programmeinclude both directly relevant and supporting documentation. They mustpermit traceability of the measurements and the dose assessment. Directlyrelevant information includes details of the individual, bioassay data,workplace monitoring data, the purpose of the monitoring and doseassessments.

Details of an individual include: a unique identifier, which may includethe name of his or her employer; his or her occupation; the radionuclides towhich he or she might be exposed; specific workplace locations and tasks; his orher work schedule; and his or her employment history, including, wherenecessary, different roles within organizations. In some exposure situations, forexample in mines, whether or not the individual smokes can influence theassessment and might therefore be usefully recorded.

35

Information to be included in records for in vivo bioassay data is given insection 5.2 of Ref. [7], for in vitro data in section 8.1 of Ref. [8] and forworkplace monitoring in section 8.5 of Ref. [6].

Dose assessment records include: computed results such as the activityconcentration in air; body activity contents or daily excretion rates and theirstatistical analyses; and computed intake values and the biokinetic models fromwhich they were derived. Dose assessment records for each confirmed intakeinclude: the intake pattern assumed; the intake route, with information onwhether it is known or assumed; the chemical form of the nuclide, withinformation on whether it is known or assumed; the particle size, withinformation on whether it is known or assumed; the classification for gases andvapours; the absorption type assumed; the committed effective doses; the dosecoefficients used; and the computer software or reference material used for thecomplete dose assessment.

If an estimate has been made of the dose to the embryo or foetus of apregnant worker, this estimate needs to be recorded. The use of the dose coeffi-cients in Ref. [46] is recommended.

In the case of long lived radionuclides, records need to reflect periodicreassessments of effective dose, based on further bioassay results and/orimproved methods for dose assessment.

Supporting documentation includes the training and qualification recordsof dose assessors, quality assurance (QA) procedures and quality control datasuch as background trends and detector efficiency.

5.3. REPORTING INFORMATION TO MANAGEMENT

The procedures and levels to be used for reporting individual doseassessment results need to be clearly specified by the management orregulatory bodies. Information reported to management needs to be clearlyidentifiable and understandable. Management may set reporting levels on suchparameters as committed effective dose, or estimated intake of activity, thatwill identify results that are to be reported, usually within specified timeperiods. Further details can be found in Ref. [6].

36

6. QUALITY ASSURANCE

6.1. GENERAL CONSIDERATIONS

The maintenance of the effectiveness of any radiation protectionprogramme relies on the ability of those in charge of implementing its variouscomponents to adopt a QA programme. General guidance on QA require-ments relating to occupational exposure are given in the BSS [2] and inRef. [47]. Additional guidance is given in Ref. [5] and in Refs [48, 49]. Thefollowing deals specifically with issues relating to the assessment of exposurefrom intakes of radionuclides.

While formal QA procedures can be applied to good effect in labora-tories carrying out measurements on individuals or on biological samples, it isdifficult to recommend similarly strict rules and procedures for assessments.The scope for subjective decisions based on experience and knowledge isinevitably much wider for the assessment stage than for the measurementstage.

A system needs to be established to provide a quality indicator of theoverall internal dosimetry service performance. Assessments of internal doseare complicated and intercomparison exercises have shown that evencompetent laboratories can arrive at very different estimates of dose given thesame original data [50]. This underlines the need for caution in internal doseassessments and the provision of suitable QA procedures where possible.

6.2. DOCUMENTATION

The QA programme related to internal exposure assessment must bethoroughly documented. A QA plan needs to be prepared that containsgeneral instructions on implementing the programme and on the various stepsin its operation. Written procedures describe every step to assess internal dosesfrom bioassay data and from workplace monitoring data, including estimates ofthe minimal detectable activities for the measurement techniques employed,and possible ‘missed doses’ for the monitoring intervals used. The proceduresalso have to contain data on quality control requirements, as, for example,national and international intercomparison exercises and training records.Documentation has to include the physical and chemical characteristics of theradionuclides present in the workplace, the methods for calculating internaldoses for specific radionuclides, the mixtures of radionuclides and materials ofdiffering chemical characteristics, the type and frequency of monitoring,

37

monitoring equipment, background trends, the selection of workers formonitoring, and the recording and reporting practices for internal doseassessment. The biokinetic and dosimetric models and the computational codesused for dose assessment need to be well defined. The methods to identifybioassay results above background values and the method to account for theportion of a bioassay result that may be due to prior intakes are to be includedin the documentation. Quality control procedures document the use of controlcharts and other methods for tracking every step for dose assessment, andcontain instructions for reporting and correcting deviations, as well as fortaking account of changes in operation. It is also necessary to prepareprocedures for documenting and reporting results. Likewise, procedures forrecord preparation, maintenance and archiving will be needed. The documen-tation provides sufficient information for an auditor to trace the operationfrom start to finish and to assess its validity.

39

Appendix I

BASIC DATA FOR INTERNAL DOSE ASSESSMENTS

This appendix provides basic data for the assessment of intakes ofradionuclides.

TABLE 1. RADIATION WEIGHTING FACTORS IN THE BSS [2]

Type and energy range of radiationRadiation weighting factor,

wR

Photons, all energiesElectrons and muons, all energiesa

Neutrons, energy:<10 keV10 keV–100 keV>100 keV–2 MeV>2 MeV–20 MeV>20 MeV

Protons, other than recoil protons, energy >2 MeVAlpha particles, fission fragments, heavy nuclei

11

5102010

55

20

a Excluding Auger electrons emitted from radionuclides bound to DNA, for whichspecial microdosimetric considerations apply.

40

TABLE 2. TISSUE WEIGHTING FACTORSIN THE BSS [2]a, b

Tissue or organ Tissue weighting factor, wT

GonadsRed bone marrowColonc

Lungd

StomachBladderBreastLiverOesophagusThyroidSkinBone surfaceRemaindere

0.200.120.120.120.120.050.050.050.050.050.010.010.05

a Values of wT originally from Ref. [1].b The values have been developed for a reference popula-

tion of equal numbers of both sexes and a wide range ofages. In the definition of effective dose they apply toworkers, to the whole population and to either sex [1].

c Doses calculated as mass weighted average to upper andlower large intestine:Hcolon = 0.57HULI + 0.43HLLI [17]

d Thoracic regions (BB, bb, AI and LNTH) of the respira-tory tract.

e For the purposes of calculation, the remainder iscomposed of the adrenal glands, brain, ET regions ofthe respiratory tract, small intestine, kidneys, muscle,pancreas, spleen, thymus and uterus. In those cases inwhich the most exposed remainder tissue receives thehighest committed equivalent dose of all organs, aweighting factor of 0.025 is to be applied to that tissue ororgan and a weighting factor of 0.025 to the mass weightedaverage dose in the rest of the remainder as defined here[18].

41

TABLE 3. COMMITTED EFFECTIVE DOSE PER UNIT INTAKE(DOSE COEFFICIENT) BY INHALATION, BY INGESTION ANDTHROUGH DIRECT INTAKE TO BLOOD FOR SELECTEDRADIONUCLIDES

Inhalation Ingestiona Injectionb

Type/formc

e(g)inh (Sv/Bq)

f1e(g)ing

(Sv/Bq)f1

e(g)inj (Sv/Bq)AMAD

= 1 µmAMAD= 5 µm

3H HTOd,e

OBTd,e

Gasd

1.8 × 10–11

4.1 × 10–11

1.8 × 10–15

———

11—

1.8 × 10–11

4.2 × 10–11

—

———

1.8 × 10–11

——

32P FM

8.0 × 10–10

3.2 × 10–91.1 × 10–9

2.9 × 10–90.8—

2.3 × 10–10

———

2.2 × 10–9

—55Fe F

M7.7 × 10–10

3.7 × 10–10 9.2 × 10–10

3.3 × 10–100.1—

3.3 × 10–10

—0.1—

3.0 × 10–9

—59Fe F

M2.2 × 10–9

3.5 × 10–93.0 × 10–9

3.2 × 10–90.1—

1.8 × 10–9

—0.1—

8.4 × 10–9

—60Co M

S9.6 × 10–9

2.9 × 10–8 7.1 × 10–9

1.7 × 10–80.10.05

3.4 × 10–9

2.5 × 10–9——

1.9 × 10–8

—67Ga F

M6.8 × 10–11

2.3 × 10–101.1 × 10–10

2.8 × 10–100.001—

1.9 × 10–10

———

1.2 × 10–10

—85Sr F

S3.9 × 10–10

7.7 × 10–105.6 × 10–10

6.4 × 10–100.30.01

5.6 × 10–10

3.3 × 10–10——

1.1 × 10–9

—89Sr F

S1.0 × 10–9

7.5 × 10–9 1.4 × 10–9

5.6 × 10–90.30.01

2.6 × 10–9

2.3 × 10–9——

3.1 × 10–9

—90Sr F

S2.4 × 10–8

1.5 × 10–7 3.0 × 10–8

7.7 × 10–80.30.01

2.8 × 10–8

2.7 × 10–9——

8.8 × 10–8

—95Zr F

MS

2.5 × 10–9

4.5 × 10–9

5.5 × 10–9

3.0 × 10–9

3.6 × 10–9

4.2 × 10–9

0.002——

8.8 × 10–10

——

———

1.0 × 10–8

——

95Nb MS

1.4 × 10–9

1.6 × 10–9 1.3 × 10–9

1.3 × 10–90.01—

5.8 × 10–10

———

2.1 × 10–9

—99Tc F

M2.9 × 10–10

3.9 × 10–94.0 × 10–10

3.2 × 10–90.8—

7.8 × 10–10

———

8.7 × 10–10

—99mTc F

M1.2 × 10–11

1.9 × 10–11 2.0 × 10–11

2.9 × 10–110.8—

2.2 × 10–11

———

1.9 × 10–11

—106Ru F

MS

8.0 × 10–9

2.6 × 10–8

6.2 × 10–8

9.8 × 10–9

1.7 × 10–8

3.5 × 10–8

0.05——

7.0 × 10–9

——

———

3.0 × 10–8

——

42

125Sb FM

1.4 × 10–9

4.5 × 10–91.7 × 10–9

3.3 × 10–90.1—

1.1 × 10–9

———

5.4 × 10–9

—123I F

Vc7.6 × 10–11

2.1 × 10–101.1 × 10–10

—1.0—

12.1 × 10–10

———

2.2 × 10–10

—124I F

Vc4.5 × 10–9

1.2 × 10–86.3 × 10–9

—1.0—

1.3 × 10–8

———

1.3 × 10–8

—125I F

Vc5.3 × 10–9

1.4 × 10–87.3 × 10–9

—1.0—

1.5 × 10–8

———

1.5 × 10–8

—131I F

Vc7.6 × 10–9

2.0 × 10–81.1 × 10–8

—1.0—

2.2 × 10–8

———

2.2 × 10–8

—134Cs F 6.8 × 10–9 9.6 × 10–9 1.0 1.9 × 10–8 — 1.9 × 10–8

137Cs F 4.8 × 10–9 6.7 × 10–9 1.0 1.3 × 10–8 — 1.4 × 10–8

144Ce MS

3.4 × 10–8

4.9 × 10–82.3 × 10–8

2.9 × 10–85 × 10–4

—5.2 × 10–9

———

1.7 × 10–7

—153Gd F

M2.1 × 10–9

1.9 × 10–92.5 × 10–9

1.4 × 10–95 × 10–4

—

12.7 × 10–10

———

8.6 × 10–9

—201Tl F 4.7 × 10–11 7.6 × 10–11 1.0 19.5 × 10–11 — 8.7 × 10–11

210Pb F 8.9 × 10–7 1.1 × 10–6 0.2 6.8 × 10–7 0.2 3.5 × 10–6

210Po FM

6.0 × 10–7

3.0 × 10–67.1 × 10–7

2.2 × 10–60.1—

2.4 × 10–7

———

2.4 × 10–6

—226Ra M 3.2 × 10–6 2.2 × 10–6 0.2 2.8 × 10–7 — 1.4 × 10–6

228Ra M 2.6 × 10–6 1.7 × 10–6 0.2 6.7 × 10–7 — 3.4 × 10–6

228Th MS

3.1 × 10–5

3.9 × 10–52.3 × 10–5

3.2 × 10–55 × 10–4

2 × 10–47.0 × 10–8

3.5 × 10–85 × 10–4

—1.2 × 10–4

—232Th M

S4.2 × 10–5

2.3 × 10–52.9 × 10–5

1.2 × 10–55 × 10–4

2 × 10–42.2 × 10–7

9.2 × 10–85 × 10–4

—4.5 × 10–4

—234U F

MS

5.5 × 10–7

3.1 × 10–6

8.5 × 10–6

6.4 × 10–7

2.1 × 10–6

6.8 × 10–6

0.020.002—

4.9 × 10–8

8.3 × 10–9

—

———

2.3 × 10–6

——

TABLE 3. COMMITTED EFFECTIVE DOSE PER UNIT INTAKE(DOSE COEFFICIENT) BY INHALATION, BY INGESTION ANDTHROUGH DIRECT INTAKE TO BLOOD FOR SELECTEDRADIONUCLIDES (cont.)

Inhalation Ingestiona Injectionb

Type/formc

e(g)inh (Sv/Bq)

f1e(g)ing

(Sv/Bq)f1

e(g)inj (Sv/Bq)AMAD

= 1 µmAMAD= 5 µm

43

235U FMS

5.1 × 10–7

2.8 × 10–6

7.7 × 10–6

6.0 × 10–7

1.8 × 10–6

6.1 × 10–6

0.020.002—

4.6 × 10–8

8.3 × 10–9

—

———

2.1 × 10–6

——

238U FMS

4.9 × 10–7

2.6 × 10–6

7.3 × 10–6

5.8 × 10–7

1.6 × 10–6

5.7 × 10–6

0.020.002—

4.4 × 10–8

7.6 × 10–9

—

———

2.1 × 10–6

——

237Np M 2.1 × 10–5 1.5 × 10–5 5 × 10–4 1.1 × 10–7 5 × 10–4 2.1 × 10–4

239Np M 9.0 × 10–10 1.1 × 10–9 5 × 10–4 8.0 × 10–10 5 × 10–4 3.8 × 10–10

238Pu MS

4.3 × 10–5

1.5 × 10–5

—

3.0 × 10–5

1.1 × 10–5

—

5 × 10–4

1 × 10–5

1 × 10–4

2.3 × 10–7

8.8 × 10–9

4.9 × 10–8

5 × 10–4

——

4.5 × 10–4

——

239Pu MS

4.7 × 10–5

1.5 × 10–5

—

3.2 × 10–5

8.3 × 10–6

—

5 × 10–4

1 × 10–5

1 × 10–4

2.5 × 10–7

9.0 × 10–9

5.3 × 10–8

5 × 10–4

——

4.9 × 10–4

——

240Pu MS

4.7 × 10–5

1.5 × 10–5

—

3.2 × 10–5

8.3 × 10–6

—

5 × 10–4

1 × 10–5

1 × 10–4

2.5 × 10–7

9.0 × 10–9

5.3 × 10–8

5 × 10–4

——

4.9 × 10–4

——

241Pu MS

8.5 × 10–7

1.6 × 10–7

—

5.8 × 10–7

8.4 × 10–8

—

5 × 10–4

1 × 10–5

1 × 10–4

4.7 × 10–9

11.1 × 10–10

9.6 × 10–10

5 × 10–4

——

9.5 × 10–6

——

241Am M 3.9 × 10–5 2.7 × 10–5 5 × 10–4 2.0 × 10–7 5 × 10–4 4.0 × 10–4

242Cm M 4.8 × 10–6 3.7 × 10–6 5 × 10–4 1.2 × 10–8 5 × 10–4 1.4 × 10–5

244Cm M 2.5 × 10–5 1.7 × 10–5 5 × 10–4 1.2 × 10–7 5 × 10–4 2.4 × 10–4

a f1 values given here apply only to ingestion, not to the inhalation doses in the adjacentcolumn.

b Direct intake into blood. For most of these cases an f1 value is not relevant. However,in some cases, such as for plutonium, the model involves recycling of material throughthe small intestine. In such cases the f1 value is used and is therefore given in the table.

c For lung absorption types, see Section 2.2.1.d For inhalation of gases and vapours, the AMAD does not apply for this form.e HTO: tritiated water; OBT: organically bound tritium.

TABLE 3. COMMITTED EFFECTIVE DOSE PER UNIT INTAKE(DOSE COEFFICIENT) BY INHALATION, BY INGESTION ANDTHROUGH DIRECT INTAKE TO BLOOD FOR SELECTEDRADIONUCLIDES (cont.)

Inhalation Ingestiona Injectionb

Type/formc

e(g)inh (Sv/Bq)

f1e(g)ing

(Sv/Bq)f1

e(g)inj (Sv/Bq)AMAD

= 1 µmAMAD= 5 µm

44

TABLE 4. DERIVED AIR CONCENTRATIONSa (DACs) FORSELECTED RADIONUCLIDES

Type/formbDAC (Bq/m3)

AMAD = 1 µm AMAD = 5 µm Gas/vapour

3H HTOc

OBTGas

———

———

5 × 105

2 × 105

5 × 109

32P FM

1 × 104

3 × 1038 × 103

3 × 103——

55Fe FM

1 × 104

2 × 1049 × 103

3 × 104——

59Fe FM

4 × 103

2 × 1033 × 103

3 × 103——

60Co MS

9 × 102

3 × 1021 × 103

5 × 102——

67Ga FM

1 × 105

4 × 1048 × 104

3 × 104——

85Sr FS

2 × 104

1 × 1041 × 104

1 × 104——

89Sr FS

8 × 103

1 × 1036 × 103

1 × 103——

90Sr FS

3 × 102

6 × 101 3 × 102

1 × 102——

95Zr FMS

3 × 103

2 × 103

2 × 103

3 × 103

2 × 103

2 × 103

———

95Nb MS

6 × 103

5 × 1036 × 103

6 × 103——

99Tc FM

3 × 104

2 × 1032 × 104

3 × 103——

99mTc FM

7 × 105

4 × 1054 × 105

3 × 105——

106Ru FMS

1 × 103

3 × 102

1 × 102

9 × 102

5 × 102

2 × 102

———

125Sb FM

6 × 103

2 × 1035 × 103

3 × 103——

45

123I FV

1 × 105

—8 × 104

——

4 × 104

124I FV

2 × 103

—1 × 103

——

7 × 102

125I FV

2 × 103

—1 × 103

——

6 × 102

131I FV

1 × 103

—8 × 102

——

4 × 102

134Cs F 1 × 103 9 × 102 —137Cs F 2 × 103 1 × 103 —144Ce M

S2 × 102

2 × 1024 × 102

3 × 102——

153Gd FM

4 × 103

4 × 1033 × 103

6 × 103——

201Tl F 2 × 105 1 × 105 —210Pb F 9 × 100 8 × 100 —210Po F

M1 × 101

3 × 1001 × 101

4 × 100——

226Ra M 3 × 100 4 × 100 —228Ra M 3 × 100 5 × 100 —228Th M

S3 × 10–1

2 × 10–14 × 10–1

3 × 10–1——

232Th MS

2 × 10–1

4 × 10–13 × 10–1

7 × 10–1——

234U FMS

2 × 101

3 × 100

1 × 100

1 × 101

4 × 100

1 × 100

———

235U FMS

2 × 101

3 × 100

1 × 100

1 × 101

5 × 100

1 × 100

———

238U FMS

2 × 101

3 × 100

1 × 100

1 × 101

5 × 100

1 × 100

———

TABLE 4. DERIVED AIR CONCENTRATIONSa (DACs) FORSELECTED RADIONUCLIDES (cont.)

Type/formbDAC (Bq/m3)

AMAD = 1 µm AMAD = 5 µm Gas/vapour

46

237Np M 4 × 10–1 6 × 10–1 —239Np M 9 × 103 8 × 103 —238Pu M

S2 × 10–1

6 × 10–13 × 10–1

8 × 10–1——

239Pu MS

2 × 10–1

6 × 10–13 × 10–1

1 × 100——

240Pu MS

2 × 10–1

6 × 10–13 × 10–1

1 × 100——

241Pu MS

1 × 101

5 × 1011 × 101

1 × 102——

241Am M 2 × 10–1 3 × 10–1 —242Cm M 2 × 100 1 × 100 —244Cm M 3 × 10–1 5 × 10–1 —

a Calculated assuming an average breathing rate of 1.2 m3/h, 2000 h worked annually anda dose limit of 20 mSv.

b For lung absorption types, see Section 2.2.1.c The DAC does not allow for absorption through intact skin.

TABLE 4. DERIVED AIR CONCENTRATIONSa (DACs) FORSELECTED RADIONUCLIDES (cont.)

Type/formbDAC (Bq/m3)

AMAD = 1 µm AMAD = 5 µm Gas/vapour

47

Appendix II

BIOKINETIC MODELS FOR SELECTED ELEMENTS AND RADIONUCLIDES

II.1. HYDROGEN

Absorption types, f1 values and chemical forms for hydrogen are given inTable 5. They are taken from the tables in Schedule II of the BSS [2] and areconsistent with those given in Refs [3, 19].

The biokinetic model adopted here in this report, and as adopted in theBSS [2], is taken from Ref. [16]. For HTO it is assumed that 97% of the activityequilibrates with body water and is retained with a biological half-life of tendays. The remaining 3% is assumed to be incorporated into organic moleculesand retained with a biological half-life of 40 days. For organically bound tritium(OBT), 50% of the activity is taken to be retained with the ten day biologicalhalf-life of water and 50% with the 40 day biological half-life of organic carbon.

For tritium gas (HT) it is assumed that 0.01% of the inhaled HT isabsorbed and converted to HTO, while for tritiated methane it is assumed that1% is metabolized and behaves as HTO [19].

The accompanying CD gives predicted fractions of intake at various timesafter a single acute intake of tritium. For HTO the concentration in urine isgiven. For OBT, total body retention and daily urine excretion are given. Dailyfaecal excretion is also given for ingestion of OBT.

TABLE 5. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR HYDROGEN

f1 Intake

Ingestion 1.0 Ingestion of HTO or OBT

Inhalation, Type V, SR-1 a Inhalation of tritium gas and tritiated methane

Inhalation, Type V, SR-2 a Inhalation of HTO and organic compounds

a Not applicable, since all activity deposited in the respiratory tract is instantaneouslyabsorbed.

48

II.2. PHOSPHORUS

Absorption types, f1 values and chemical forms for phosphorus are givenin Table 6. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [14]. Following entry into the transfer compartment, 30% ofthe activity is taken to be deposited in bone (32P on bone surfaces), where it isretained with a retention half-life of 1500 days. Fifty-five per cent of the activityis deposited in other tissues; of this, 40% is retained with a biological half-life of19 days and 15% with a biological half-life of two days. The remaining 15% ofthe activity is taken to be excreted promptly (with a biological half-life of0.5 day).

For activity lost to excretion from systemic compartments, 50% isassumed to be lost to urine and 50% to faeces [3]. Note that the predictedvalues of daily excretion in urine and faeces (see Section 2.4) are assumed andtherefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 32P: daily urinary and faecalexcretion; total body retention; retention in the lungs for intakes of Type Mmaterial by inhalation; and retention in the skeleton.

TABLE 6. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR PHOSPHORUS

f1 Compound

Ingestion 0.8 All compounds

Inhalation, Type F 0.8 All unspecified compounds

Inhalation, Type M 0.8 Some phosphates: determined by combining cation

49

II.3. IRON

Absorption types, f1 values and chemical forms for iron are given inTable 7. They are taken from the BSS [2] and are consistent with those given inRef. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [18]. Following entry into the transfer compartment, most ironis transported to the red bone marrow, incorporated into haemoglobin in newlyformed erythrocytes and re-released to the circulation. Smaller amounts of ironare stored in other tissues, principally the liver. Iron from senescent red bloodcells is transferred mainly to the red bone marrow, liver and spleen. Losses ofiron from the body are largely due to exfoliation of cells from the skin and theGI tract, with smaller amounts in sweat, bile and urine.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 55Fe or 59Fe: daily urinary and faecalexcretion; total body retention; retention in the lungs for intakes of Type Mmaterial by inhalation; and retention in the liver.

TABLE 7. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR IRON

f1 Compound

Ingestion 0.1 All compounds

Inhalation, Type F 0.1 All unspecified compounds

Inhalation, Type M 0.1 Oxides, hydroxides and halides

50

II.4. COBALT

Absorption types, f1 values and chemical forms for cobalt are given inTable 8. They are taken from the BSS [2] and are consistent with those given inRef. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. Following entry into the transfer compartment, 50% ofthe cobalt is rapidly excreted with a biological half-life of 0.5 day, 5% is takenup by the liver and 45% is uniformly distributed in all other tissues. Fractions of0.6, 0.2 and 0.2 are assumed to be lost from the liver and other tissues, withbiological half-lives of 6, 60 and 800 days, respectively. For activity lost toexcretion from systemic compartments, 86% is assumed to be lost to urine and14% to faeces. Note that the predicted values of daily excretion in urine andfaeces (see Section 2.4) are assumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 60Co: daily urinary and faecalexcretion; total body retention; retention in the lungs for intakes by inhalation;and retention in the liver.

TABLE 8. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR COBALT

f1 Compound

Ingestion 0.1 All unspecified compounds

Ingestion 0.05 Oxides, hydroxides and inorganic compounds

Inhalation, Type M 0.1 All unspecified compounds

Inhalation, Type S 0.05 Oxides, hydroxides, halides and nitrates

51

II.5. GALLIUM

Absorption types, f1 values and chemical forms for gallium are given inTable 9. They are taken from the BSS [2] and are consistent with those given inRef. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [40]. Following entry into the transfer compartment, 30% isdeposited on bone surfaces, 9% in the liver, 1% in the spleen and 60% in allother tissues. Gallium is retained in all organs and tissues with biological half-lives of one day (30%) and 50 days (70%). For activity lost to excretion fromsystemic compartments, 50% is assumed to be lost to urine and 50% to faeces[3]. Note that the predicted values of daily excretion in urine and faeces (seeSection 2.4) are assumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 67Ga: total body retention; retentionin the lungs for intakes of Type M materials by inhalation; retention in the liver;and retention in the skeleton.

TABLE 9. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR GALLIUM

f1 Compound

Ingestion 0.001 All compounds

Inhalation, Type F 0.001 All unspecified compounds

Inhalation, Type M 0.001 Oxides, hydroxides, carbides, halides and nitrates

52

II.6. STRONTIUM

Absorption types, f1 values and chemical forms for strontium are given inTable 10. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. This model describes in detail the kinetics of alkalineearth elements in bone, which is the main site of deposition and retention, andconsiders also retention in soft tissues and routes of excretion. It takes accountof initial uptake on to bone surfaces, transfer from the surface to bone volumeand recycling from bone and soft tissues to blood. It also describes theexcretion routes for which no constant ratio is used.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 85Sr, 89Sr and 90Sr: daily urinary andfaecal excretion; total body retention; retention in the lungs for intakes ofType S materials by inhalation; and retention in the skeleton.

TABLE 10. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR STRONTIUM

f1 Compound

Ingestion 0.3 All unspecified compounds

Ingestion 0.01 Strontium titanate (SrTiO3)

Inhalation, Type F 0.3 All unspecified compounds

Inhalation, Type S 0.01 Strontium titanate (SrTiO3)

53

II.7. ZIRCONIUM

Absorption types, f1 values and chemical forms for zirconium are given inTable 11. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. Following entry into the transfer compartment, 50% ofsystemic zirconium is retained on bone surfaces with a biological half-life of10 000 days (relating to the rate of bone remodelling), and the other 50% isdistributed throughout all other tissues and is retained with a biological half-life of seven days. For activity lost to excretion from systemic compartments,83% is assumed to be lost to urine and 17% to faeces. Note that the predictedvalues of daily excretion in urine and faeces (see Section 2.4) are assumed andtherefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 95Zr: daily urinary and faecalexcretion; total body retention; retention in the lungs for intakes of Type Mmaterial by inhalation; and retention in the skeleton.

TABLE 11. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR ZIRCONIUM

f1 Compound

Ingestion 0.002 All compounds

Inhalation, Type F 0.002 All unspecified compounds

Inhalation, Type M 0.002 Oxides, hydroxides, halides and nitrates

Inhalation, Type S 0.002 Zirconium carbide

54

II.8. NIOBIUM

Absorption types, f1 values and chemical forms for niobium are given inTable 12. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [16]. Following entry into the transfer compartment, 0.4 isdeposited in mineral bone, 0.2 in the liver, 0.03 in the kidneys and 0.37 in allother tissues. The retention is described by a two component exponentialfunction for all tissues and organs, with biological half-lives of six days (50%)and 200 days (50%). Niobium-95 in the skeleton is assumed to be distributedover bone surfaces. For activity lost to excretion from systemic compartments,83% is assumed to be lost to urine and 17% to faeces [17]. Note that thepredicted values of daily excretion in urine and faeces (see Section 2.4) areassumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 95Nb: daily urinary and faecalexcretion; total body retention; retention in the lungs for intakes by inhalation;and retention in the skeleton.

TABLE 12. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR NIOBIUM

f1 Compound

Ingestion 0.01 All compounds

Inhalation, Type M 0.01 All unspecified compounds

Inhalation, Type S 0.01 Oxides and hydroxides

55

II.9. TECHNETIUM

Absorption types, f1 values and chemical forms for technetium are givenin Table 13. They are taken from the BSS [2] and are consistent with thosegiven in Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. Following entry into the transfer compartment, 0.04 oftechnetium is taken up by the thyroid gland and retained with a biological half-life of 0.5 day. Further fractions of 0.1 and 0.03 are assumed to be translocatedto the stomach wall and liver, respectively, and the remaining fraction isassumed to be uniformly distributed in all other tissues. Biological half-lives forthe retention of technetium in all tissues other than the thyroid are taken to be1.6, 3.7 and 22 days, applying to fractions of 0.75, 0.2 and 0.05, respectively. Thebiological half-life in blood is assumed to be 0.02 day. For activity lost toexcretion from systemic compartments, 50% is assumed to be lost to urine and50% to faeces. Note that the predicted values of daily excretion in urine andfaeces (see Section 2.4) are assumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 99mTc and 99Tc: daily urinary andfaecal excretion; total body retention; and retention in the lungs for intakes ofType M material by inhalation. Data are not given for specific tissues besidesthe lungs, since the tissue activity concentrations do not significantly exceed theaverage activity distribution.

TABLE 13. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR TECHNETIUM

f1 Compound

Ingestion 0.8 All compounds

Inhalation, Type F 0.8 All unspecified compounds

Inhalation, Type M 0.8 Oxides, hydroxides, halides and nitrates

56

II.10. RUTHENIUM

Absorption types, f1 values and chemical forms for ruthenium are given inTable 14. They are taken from the BSS [2] and are consistent with those givenin Refs [3, 19].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [16]. For ruthenium absorbed to body fluids, data have shownthat the subsequent tissue distribution is fairly uniform. A model using a threeterm retention expression is recommended: 35% of activity is retained with abiological half-life of eight days, 30% with 35 days and 20% with 1000 days. Thebiological half-life in body fluids is taken to be 0.3 day, and 15% of systemicactivity is assumed to be excreted directly. For activity lost to excretion fromsystemic compartments, 80% is assumed to be lost to urine and 20% to faeces.Note that the predicted values of daily excretion in urine and faeces (seeSection 2.4) are assumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 106Ru: daily urinary and faecalexcretion; total body retention; and retention in the lungs for intakes byinhalation of Type M and Type S material. Data are not given for specifictissues, besides the lungs, since the tissue activity concentrations do not signifi-cantly exceed the average activity distribution.

TABLE 14. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR RUTHENIUM

f1 Compound

Ingestion 0.05 All compounds

Inhalation, Type F 0.05 Unspecified compounds

Inhalation, Type M 0.05 Halides

Inhalation, Type S 0.05 Oxides and hydroxides

Inhalation 0.05 Tetroxide

57

II.11. ANTIMONY

Absorption types, f1 values and chemical forms for antimony are given inTable 15. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [18]. From that part of antimony entering the circulation, afraction of 0.2 is rapidly excreted, 0.4 is taken up by bone surfaces, 0.05 by theliver and the remaining fraction of 0.35 is uniformly distributed throughout allother organs. For all tissues, fractions of 0.85, 0.1 and 0.05 are assumed to beretained with biological half-lives of 5, 100 and 5000 days, respectively. Foractivity lost to excretion from systemic compartments, 80% is assumed to belost to urine and 20% to faeces. Note that the predicted values of dailyexcretion in urine and faeces (see Section 2.4) are assumed and therefore willvary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 125Sb: daily urinary and faecalexcretion; total body retention; retention in the lungs for intakes of Type Mmaterial by inhalation; and retention in the skeleton.

TABLE 15. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR ANTIMONY

f1 Compound

Ingestion 0.1 All compounds

Inhalation, Type F 0.1 Unspecified compounds

Inhalation, Type M 0.01 Oxides, hydroxides, halides, sulphides, sulphates and nitrates

58

II.12. IODINE

Absorption types, f1 values and chemical forms for iodine are given inTable 16. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [16]. It is assumed that, of the iodine reaching the blood, afraction of 0.3 is accumulated in the thyroid gland and 0.7 is excreted directly inurine. The biological half-life in blood is taken to be 0.25 day. Iodine incorpo-rated into thyroid hormones leaves the thyroid gland with a biological half-lifeof 80 days and enters other tissues, where it is retained with a biological half-lifeof 12 days. Most iodine (80%) is subsequently released to the blood and isavailable in the circulation for uptake by the thyroid gland and urinaryexcretion; the remainder (20%) is excreted in faeces in organic form.

The biokinetic model for iodine assumes that 0.3 is taken up by thethyroid and the remainder is excreted in urine. In fact, there are relatively largevariations, depending on many parameters such as the stable iodine content incommon food and thyroid dysfunctions; for example, current uptake values fora European euthyroid adult are in the range 0.20–0.25. However, in countrieswith iodine deficiency in food, this value is considerably higher. Pathologicalstates of the thyroid may result in uptake values of 0–0.05 (blocked thyroid) tomore than 0.5. When such cases are suspected, then individual values need tobe introduced in the dose calculation, especially for accidental exposuresleading to significant doses, for which a precise assessment is needed.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 123I, 124I, 125I and 131I: daily urinaryand faecal excretion; total body retention; and retention in the thyroid.

TABLE 16. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR IODINE

f1 Compound

Ingestion 1.0 All compounds

Inhalation, Type F 1.0 All particulate compounds

Inhalation, Type F, SR-1 1.0 Elemental iodine

Inhalation, Type V, SR-1 a Methyl iodide

a Not applicable, since all activity deposited in the respiratory tract is instantaneouslyabsorbed.

59

II.13. CAESIUM

Absorption types, f1 values and chemical forms for caesium are given inTable 17. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [16]. Following entry into the transfer compartment, caesium istaken to be distributed uniformly throughout all body tissues; 10% of theactivity is assumed to be retained with a biological half-life of two days and90% with 110 days. For females, however, the biological half-life for the longterm component is significantly less than for males [43, 46]. There is alsoevidence that in some countries the mean biological half-life of caesium inadult males is less than 110 days [51, 52]. Additionally, there is information thata small part of activity is retained with a longer biological half-life, of about500 days [43]. For activity lost to excretion from systemic compartments, 80% isassumed to be lost to urine and 20% to faeces, as per Ref. [17]. Note that thepredicted values of daily excretion in urine and faeces (see Section 2.4) areassumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 134Cs and 137Cs: daily urinary andfaecal excretion; and total body retention. No data are given for specific tissues,because the activity concentration does not significantly exceed the averageactivity distribution for any organ or tissue.

TABLE 17. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR CAESIUM

f1 Compound

Ingestion 1.0 All compounds

Inhalation, Type F 1.0 All compounds

60

II.14. CERIUM

Absorption types, f1 values and chemical forms for cerium are given inTable 18. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [16]. Following entry into the transfer compartment, cerium istaken to be distributed in the skeleton (bone surfaces: 30%), the liver (50%)and other tissues (20%). The retention half-life is taken to be 3500 days in alltissues. For activity lost to excretion from systemic compartments, 10% isassumed to be lost to urine and 90% to faeces, as per Ref. [17]. Note that thepredicted values of daily excretion in urine and faeces (see Section 2.4) areassumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 144Ce: daily urinary and faecalexcretion; total body retention; lung retention for inhalation; retention in theskeleton; and retention in the liver.

TABLE 18. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR CERIUM

f1 Compound

Ingestion 0.0005 All compounds

Inhalation, Type M 0.0005 All unspecified compounds

Inhalation, Type S 0.0005 Oxides, hydroxides and fluorides

61

II.15. GADOLINIUM

Absorption types, f1 values and chemical forms for gadolinium are givenin Table 19. They are taken from the BSS [2] and are consistent with thosegiven in Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [40]. Following entry into the transfer compartment,gadolinium is taken to be distributed to the kidneys (3%), liver (30%) andskeleton (45%), with 22% of material being promptly excreted. Gadolinium inthe liver and skeleton is assumed to be retained with a biological half-life of3500 days, while that in the kidneys is taken to be retained with a biologicalhalf-life of ten days. For activity lost to excretion, 50% is assumed to be lost tourine and 50% to faeces, following Ref. [3]. Note that the predicted values ofdaily excretion in urine and faeces (see Section 2.4) are assumed and thereforewill vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 153Gd: total body retention;retention in the lungs for intakes by inhalation; retention in the skeleton; andretention in the liver.

TABLE 19. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR GADOLINIUM

f1 Compound

Ingestion 0.0005 All compounds

Inhalation, Type F 0.0005 Unspecified compounds

Inhalation, Type M 0.0005 Oxides, hydroxides and fluorides

62

II.16. THALLIUM

Absorption types, f1 values and chemical forms for thallium are given inTable 20. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [40]. Following entry into the transfer compartment, thalliumis taken to be distributed instantaneously within the kidneys (3%) and all otherorgans (97%). Thallium in all tissues is assumed to be retained with a biologicalhalf-life of ten days. For activity lost to excretion from systemic compartments,50% is assumed to be lost to urine and 50% to faeces, as per Ref. [3]. Note thatthe predicted values of daily excretion in urine and faeces (see Section 2.4) areassumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 201Tl: total body retention.

TABLE 20. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR THALLIUM

f1 Compound

Ingestion 1.0 All compounds

Inhalation, Type F 1.0 All compounds

63

II.17. LEAD

Absorption types, f1 values and chemical forms for lead are given inTable 21. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. The model uses the structure of the alkaline earth model[17]; it describes the kinetics of lead in bone, which is the main site ofdeposition and retention, and also considers retention in the liver, red bloodcells and other soft tissues, as well as routes of excretion. It takes account ofinitial uptake on to bone surfaces, transfer from the surface to bone volume andrecycling from bone and other tissues to plasma.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 210Pb: daily urinary and faecalexcretion; total body retention; and skeleton retention.

TABLE 21. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR LEAD

f1 Compound

Ingestion 0.2 All compounds

Inhalation, Type F 0.2 All compounds

64

II.18. POLONIUM

Absorption types, f1 values and chemical forms for polonium are given inTable 22. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. Following entry into the transfer compartment, poloniumis taken to be distributed to the liver (30%), the kidneys (10%), red bonemarrow (10%), the spleen (5%) and all other tissues (45%). The retention half-life for polonium is taken to be 50 days for all tissues. For activity lost toexcretion from systemic compartments, 33% is assumed to be lost to urine and67% to faeces. Note that the predicted values of daily excretion in urine andfaeces (see Section 2.4) are assumed and therefore will vary in practice.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 210Po: daily urinary and faecalexcretion; total body retention; lung retention for intakes of Type M materialby inhalation; and retention in the skeleton.

TABLE 22. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR POLONIUM

f1 Compound

Ingestion 0.1 All compounds

Inhalation, Type F 0.1 All unspecified compounds

Inhalation, Type M 0.1 Oxides, hydroxides and nitrates

65

II.19. RADIUM

Absorption types, f1 values and chemical forms for radium are given inTable 23. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17]. The model describes the kinetics of radium in bone, whichis the main site of deposition and retention, and also considers retention in theliver and other soft tissues, as well as routes of excretion. It takes account ofinitial uptake on to bone surfaces, transfer from the surface to bone volume andrecycling from bone and other tissues to plasma.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 226Ra and 228Ra: daily urinary andfaecal excretion; total body retention; lung retention for intakes by inhalation;and skeleton retention.

TABLE 23. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR RADIUM

f1 Compound

Ingestion 0.2 All compounds

Inhalation, Type M 0.2 All compounds

66

II.20. THORIUM

Absorption types, f1 values and chemical forms for thorium are given inTable 24. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [18] and is based on the actinide model of Ref. [17]. It takesaccount of the initial deposition in bone, the liver, gonads and other tissues, andallows for transfer of activity from bone surfaces to bone volume and marrow,recycling of activity between tissues, as well as loss by excretion.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 228Th and 232Th: daily urinary andfaecal excretion; total body retention; lung retention for intakes by inhalation;retention in the liver; and retention in the skeleton.

TABLE 24. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR THORIUM

f1 Compound

Ingestion 0.0005 Unspecified compounds

Ingestion 0.0002 Oxides and hydroxides

Inhalation, Type M 0.0005 Unspecified compounds

Inhalation, Type S 0.0002 Oxides and hydroxides

67

II.21. URANIUM

Absorption types, f1 values and chemical forms for uranium are given inTable 25. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [18] and is based on the alkaline earth model of Ref. [17]. Themodel describes in detail the kinetics of uranium in bone, which is the main siteof deposition and retention, and also considers retention in the liver, thekidneys and other soft tissues, as well as routes of excretion. It takes account ofinitial uptake on to bone surfaces, transfer from the surface to bone volume andrecycling from bone and other tissues to plasma.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 234U, 235U and 238U: daily urinary andfaecal excretion; total body retention; lung retention for intakes of Type M andType S material by inhalation; retention in the kidneys; and retention in theskeleton.

TABLE 25. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR URANIUM

f1 Compound

Ingestion 0.02 Unspecified compounds

Ingestion 0.002 Most tetravalent compounds, for example UO2, U3O8 and UF4

Inhalation, Type F 0.02 Soluble compounds, including hexavalent compounds, for example UF6, UO2F2 and UO2 (NO3)2

Inhalation, Type M 0.02 Less soluble compounds, for example UO3, UF4, UCl4 and most other hexavalent compounds

Inhalation, Type S 0.002 Highly insoluble compounds, for example UO2 and U3O8

68

II.22. NEPTUNIUM

Absorption types, f1 values and chemical forms for neptunium are givenin Table 26. They are taken from the BSS [2] and are consistent with thosegiven in Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17] and is based on the actinide model. It takes account of theinitial deposition in bone, the liver, gonads and other tissues, and allows fortransfer of activity from bone surfaces to bone volume and marrow, recycling ofactivity between tissues, as well as loss by excretion.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 237Np and 239Np: daily urinary andfaecal excretion; total body retention; lung retention for intakes by inhalation;retention in the liver; and retention in the skeleton.

TABLE 26. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR NEPTUNIUM

f1 Compound

Ingestion 0.0005 All compounds

Inhalation, Type M 0.0005 All compounds

69

II.23. PLUTONIUM

Absorption types, f1 values and chemical forms for plutonium are given inTable 27. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17] and is based on the actinide model. It takes account of theinitial deposition in bone, the liver, gonads and other tissues, and allows fortransfer of activity from bone surfaces to bone volume and marrow, recycling ofactivity between tissues, as well as loss by excretion.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 238Pu, 239Pu, 240Pu and 241Pu: dailyurinary and faecal excretion; total body retention; lung retention for intakes byinhalation; retention in the liver; and retention in the skeleton.

TABLE 27. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR PLUTONIUM

f1 Compound

Ingestion 0.0005 Unspecified compounds

Ingestion 0.0001 Nitrates

Ingestion 0.00001 Insoluble oxides

Inhalation, Type M 0.0005 Unspecified compounds

Inhalation, Type S 0.00001 Insoluble oxides

70

II.24. AMERICIUM

Absorption types, f1 values and chemical forms for americium are given inTable 28. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [17] and is based on the actinide model. It takes account of theinitial deposition in bone, the liver, gonads and other tissues, and allows fortransfer of activity from bone surfaces to bone volume and marrow, recycling ofactivity between tissues, as well as loss by excretion.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 241Am: daily urinary and faecalexcretion; total body retention; lung retention for intakes by inhalation;retention in the liver; and retention in the skeleton.

TABLE 28. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR AMERICIUM

f1 Compound

Ingestion 0.0005 All compounds

Inhalation, Type M 0.0005 All compounds

71

II.25. CURIUM

Absorption types, f1 values and chemical forms for curium are given inTable 29. They are taken from the BSS [2] and are consistent with those givenin Ref. [3].

The biokinetic model adopted here, and as adopted in the BSS [2], istaken from Ref. [19] and is identical to the americium model of Ref. [17]. Ittakes account of the initial deposition in bone, the liver, gonads and othertissues, and allows for transfer of activity from bone surfaces to bone volumeand marrow, recycling of activity between tissues, as well as loss by excretion.

The accompanying CD gives the following predicted fractions of intake atvarious times after a single acute intake of 242Cm and 244Cm: daily urinary andfaecal excretion; total body retention; lung retention for intakes by inhalation;retention in the liver; and retention in the skeleton.

TABLE 29. COMPOUNDS, ABSORPTION TYPES AND f1 VALUESFOR CURIUM

f1 Compound

Ingestion 0.0005 All compounds

Inhalation, Type M 0.0005 All compounds

72

Appendix III

RETENTION AND EXCRETION FRACTIONS FOR INTAKES OF SELECTED RADIONUCLIDES

As a primary complement to this appendix, the CD that is attached to thisreport contains tables showing the m(t) values for the retention and excretionfunctions for intakes of selected radionuclides. The following exposurepathways are reviewed: inhalation of 1 µm AMAD particles, inhalation of 5 µmAMAD particles, inhalation of gases and vapours (for some elements), andingestion and injection (i.e. direct intake to blood). If a given radionuclide hasmore than one absorption type or value for f1, tables for each are included.These values are based on biokinetic models, and resulting estimates of intakesmay be used with the dose coefficients given in Table 3 to compute the effectivedose to workers.

The predicted values of retention and excretion of activity given here (i.e.on the CD) are derived from the latest models recommended by the ICRP.Some of these models are more suitable for this application than others. Inparticular, two aspects of the tables need some explanation.

Firstly, the excretion model for tritium in Ref. [21] can be reliably appliedonly relatively soon after intake. For this reason, the tables given for tritium aretruncated at 100 days after intake.

Secondly, in the models for gallium, gadolinium and thallium, theexcretion routes have not been defined as they have for other models. Resultsfor daily urinary and faecal excretion are therefore not given for theseelements. Results for retention of these elements in regions of the body areunaffected by the excretion modelling and are therefore given.

The m(t) values are given for time since intake in days, on an expandingscale, that is t = 1, 2, 3,…, 10, 20, 30…, etc. To obtain a value for a time notlisted, a logarithmic interpolation between adjacent values is needed. If thevalues are not changing rapidly, a linear interpolation may be sufficientlyprecise.

The data files on the accompanying CD were produced and qualityassured by A.W. Phipps of the National Radiological Protection Board in theUK and D. Noßke of the Bundesamt für Strahlenschutz in Germany using thePLEIADES and DOSAGE computer codes, respectively. Both codes havebeen used for a number of years by the ICRP Task Group on Dose Calcula-tions (DOCAL) to derive dose coefficients for such documents as Refs [3, 53]and the excretion coefficients in Ref. [20]. The data files on the accompanyingCD are essentially an extension of Ref. [20].

73

Other sources of m(t) are available [54, 55], which in some areas giveadditional results to those provided here. The earlier comments in this reportregarding the limitations of the models in predicting m(t), for example at earlytimes, or where excretion models are not well established, are reiterated.

75

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[38] LEGGETT, R.W., BOUVILLE, A., ECKERMAN, K.F., Reliability of ICRP’ssystemic biokinetic models, Radiat. Prot. Dosim. 79 (1998) 335–342.

[39] NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASURE-MENTS, Evaluating the Reliability of Biokinetic and Dosimetric Models andParameters Used to Assess Individual Doses for Risk Assessment Purposes,Commentary No. 15, NCRP, Bethesda, MD (1998).

[40] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,Limits for Intakes of Radionuclides by Workers: Part 3, Publication 30, PergamonPress, Oxford and New York (1981).

[41] NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASURE-MENTS, Deposition, Retention and Dosimetry of Inhaled RadioactiveSubstances, Rep. No. 125, NCRP, Bethesda, MD (1997).

[42] BAILEY, M.R., BIRCHALL, A., Book review: Deposition, retention anddosimetry of inhaled radioactive substances, NCRP Report No. 125, Radiat. Prot.Dosim. 72 (1997) 147–151.

[43] MELO, D.R., et al., A biokinetic model for 137Cs, Health Phys. 73 (1997) 320–332.

78

[44] HURSH, J.B., SUOMEL, A., Absorption of Pb-212 from the gastrointestinal tractof man, Acta. Radiol. Ther. Phys. Biol. 7 (1968) 108–120.

[45] DOLPHIN, G.W., Dietary intakes of iodine and thyroid dosimetry, Health Phys.21 (1971) 711–712.

[46] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,Doses to the Embryo and Foetus from Intakes of Radionuclides by the Mother,Publication 88, Pergamon Press, Oxford and New York (2001).

[47] INTERNATIONAL ATOMIC ENERGY AGENCY, Quality Assurance forSafety in Nuclear Power Plants and Other Nuclear Installations, Safety SeriesNo. 50-C/SG-Q, IAEA, Vienna (1996).

[48] INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, GeneralRequirements for the Competence of Calibration and Testing Laboratories, ISO/IEC Guide 25, ISO, Geneva (1990).

[49] INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, GeneralRequirements for the Competence of Testing and Calibration Laboratories, ISO/IEC 17025, ISO, Geneva (1999).

[50] INTERNATIONAL ATOMIC ENERGY AGENCY, Intercomparison andBiokinetic Model Validation of Radionuclide Intake Assessment — Report of aCo-ordinated Research Project 1996–1998, IAEA-TECDOC-1071, IAEA,Vienna (1999).

[51] SUOMELA, M., “Elimination rate of 137Cs in individuals and in the controlgroup”, Proc. Nordic Society for Radiation Protection, Copenhagen, Institute ofRadiation Hygiene, Copenhagen (1971) 285–298.

[52] HOSONEN, E., RAHOLA, T., The biological half-life of 137Cs and 24Na in man,Ann. Clin. Res. 3 (1971) 236–240.

[53] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,Age-dependent Doses to Members of the Public from Intake of Radionuclides:Part 5, Publication 72, Pergamon Press, Oxford and New York (1996).

[54] POTTER, C.A., Intake retention fractions developed from models used in thedetermination of dose coefficients developed for the ICRP Publication 68 —particulate inhalation, Health Phys. 83 (2002) 594–789.

[55] ISHIGURE, N., Electronic look-up tables on retention and excretion of radio-nuclides as a PC based support system for internal dosimetry, Radiat. Prot.Dosim. 93 (2001) 161–165.

79

Annex I

DETERMINING INTAKE FROM SINGLE AND FROM MULTIPLE DATA MEASUREMENTS FOR DOSE ASSESSMENT

I–1. A SINGLE MEASUREMENT

For a suspected intake by inhalation of 137Cs by a male worker, themeasurement shown in Table I–1 was taken two days after the suspectedincident.

I–1.1. Solution

Since the volume of urine collected is equal to the reference daily urinaryoutput for an adult female, the measurement does not need to be adjusted. Theestimate of intake, I, is thus simply given by (see Section 3.3.3):

I = 50 kBq/0.011

= 4.5 MBq

From Table 3, the dose conversion factor for inhalation of 137Cs (5 µmAMAD) is 6.7 × 10–9 Sv/Bq, so the estimated effective dose from this intake is:

E(50) = (6.7 × 10–9 Sv/Bq)(4.5 × 106 Bq)

= 0.03 Sv, or 30 mSv

TABLE I–1. MEASUREMENT TAKEN TWO DAYS AFTERTHE SUSPECTED INCIDENT

Day Urine volume Urine activity m(t) (from Appendix III, 5 µm AMAD)

2 1400 mL 50 kBq 0.011

80

I–2. MULTIPLE DATA POINTS: SIMPLE AVERAGE

Further measurements were then made seven and ten days after theincident, as shown in Table I–2.

I–2.1. Solution

In these cases, since urine volumes are substantially lower than thereference daily urinary output for an adult male of 1.4 L1, the measurementshave to be adjusted as follows (see Section 3.3.3):

Day 7: Adjusted activity = (0.9 kBq)(1400/70) = 18 kBq

Day 10: Adjusted activity = (1.2 kBq)(1400/140) = 12 kBq

Point estimates of the intake are obtained from each of these new data inthe same manner as above.

Day 7: I = 18 kBq/0.0038 = 4.7 MBq

Day 10: I = 12 kBq/0.0026 = 4.6 MBq

Combining these with the estimate of intake obtained from themeasurement after two days gives three estimates, 4.5, 4.7 and 4.6 MBq, ofwhich the mean is 4.6 MBq.

1 This example is based on real data gathered some years ago; it therefore uses avalue for the daily urine volume taken from Ref. [I–1], which has been superseded byRef. [I–2].

TABLE I–2. MEASUREMENTS MADE SEVEN AND TEN DAYSAFTER THE INCIDENT

Day Urine volume Urine activity m(t) (from Appendix III)

7 70 mL 0.9 kBq 0.0038

10 140 mL 1.2 kBq 0.0026

81

From Table 3, the dose conversion factor for inhalation of 137Cs is6.7 × 10–9 Sv/Bq, so the estimated effective dose from this intake is:

E(50) = (6.7 × 10–9 Sv/Bq)(4.6 × 106 Bq)

= 0.03 Sv, or 30 mSv

I–3. MULTIPLE DATA POINTS: UNWEIGHTED LEAST SQUARES FIT (SECTION 3.3.4.2)

I–3.1. Solution

The least squares method may also be used to estimate an intake from thethree measurements. The relevant products, M(t)m(t) and [m(t)]2, and theirsums are given in Table I–3, along with the data.

The estimated intake using the unweighted least squares fit is:

I = 0.65/1.4 × 10–4 kBq

= 4.6 MBq

From Table 3, the dose conversion factor for inhalation of 137Cs is6.7 × 10–9 Sv/Bq, so the estimated effective dose from this intake is:

E(50) = (6.7 × 10–9 Sv/Bq)(4.6 × 106 Bq)

= 0.03 Sv, or 30 mSv

TABLE I–3. RELEVANT PRODUCTS AND DATA

Day M(t) (kBq)M(t) (from

Appendix III)M(t)m(t) [m(t)]2

2 50 0.011 0.55 1.2 × 10–4

7 18 0.0038 0.068 1.4 × 10–5

10 12 0.0026 0.031 6.8 × 10–6

Sums 0.65 1.4 × 10–4

82

In this example the three simple estimates based on single measurementswere in good agreement, indicating that the standard biokinetic model used islikely to be appropriate. Since this intake results in a rather significant effectivedose, further measurements are advisable.

REFERENCES TO ANNEX I

[I–1] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,Reference Man: Anatomical, Physiological and Metabolic Characteristics, Publi-cation 23, Pergamon Press, Oxford and New York (1975).

[I–2] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,Basic Anatomical and Physiological Data for Use in Radiological Protection:Reference Values, Publication 89, Pergamon Press, Oxford and New York (2002).

83

Annex II

DETERMINING THE TIME OF INTAKE FOR DOSE ASSESSMENT

II–1. INCIDENT

A technician was exposed to 131I vapour. A routine monitoringprogramme with a 15 day interval yielded the following results: Date:4 December; thyroid (in vivo monitoring): 710 Bq; urine (in vitro monitoring):126 Bq/24 h.

Confirmatory and investigative monitoring was carried out, with thefollowing results: Date: 6 December; thyroid (in vivo monitoring): 680 Bq;urine (in vitro monitoring): <MDA (minimum detectable activity, 1 Bq/L or1.4 Bq/24 h). Date: 8 December; thyroid (in vivo monitoring): 490 Bq; urine(in vitro monitoring): <MDA.

II–2. SOLUTION

In order to use the m(t) tables in Appendix III, it is necessary to establishthe time at which the intake occurred. In many circumstances this time is notknown a priori. Owing to the specific biokinetic behaviour of some radio-nuclides, comparing the results of different bioassay techniques may shed somelight on the time of intake.

One such example is this intake of elemental iodine: On 4 Decembermonitoring showed a ratio of 0.18 between the activities in urine over those in

TABLE II–1. PREDICTED VALUES (Bq PER Bq INTAKE)FOR INHALATION OF 131I VAPOUR

Days after intake ThyroidDaily

urinaryexcretion

Expectedurine/thyroid

ratio

1 2.30 × 10–1 5.30 × 10–1 2.30

2 2.20 × 10–1 4.30 × 10–2 0.20

3 2.00 × 10–1 2.50 × 10–3 0.0125

4 1.90 × 10–1 2.70 × 10–4 0.00142

5 1.70 × 10–1 1.70 × 10–4 0.00100

6 1.50 × 10–1 1.80 × 10–4 0.00120

84

the thyroid. The m(t) values for inhalation of 131I vapour, taken from Appen-dix III, and the expected ratios for the activities in urine and in the thyroid,are given in Table II–1.

Thus it is possible to conclude that the intake occurred on 2 December,two days before the monitoring on 4 December, since the results of theconfirmatory monitoring are compatible with this date of intake.

Four days after the intake, the amount expected in the thyroid wouldhave been:

(0.19/0.22) × 710 = 613 Bq

On the same day the amount expected in urine would have been:

(2.7 × 10–4/4.3 × 10–2) × 126 = 0.8 Bq/24 h

Six days after the intake the amount expected in the thyroid would havebeen:

(0.15/0.22) × 710 = 484 Bq

On the same day the amount expected in urine would have been:

(1.8 × 10–4/4.3 × 10–2) × 126 = 0.5 Bq/24 h

Thus one may assume an intake two days before the date of the routinemonitoring. The intake may be determined using the thyroid results. Using them(t) values given above, taken from Appendix III, shown in Table II–2, theaverage of the point estimates of the intake is 3358 Bq.

In this example the results for thyroid monitoring were used instead ofthose for urine, since they consist of a direct measurement of activity in thebody, are less time dependent and thus provide the most accurate assessment of

TABLE II–2. ESTIMATED INTAKE VALUES

Days after intakeThyroid activity

(Bq)m(t)

(from Appendix III)Point estimates

of the intake (Bq)

2 710 0.22 3227

4 680 0.19 3579

6 490 0.15 3267

85

internal contamination. In this example the urine results would produce anintake similar to the ones obtained with the thyroid data, and could have beentaken into consideration when deriving the intake. When urine samples do notproduce intake results as close to the thyroid monitoring data, in vivomonitoring results need to be used.

From Table 3, the dose conversion factor for inhalation of vapour 131I is2 × 10–8 Sv/Bq. The estimated effective dose is:

E(50) = 3358 × 2 × 10–8 = 6716 × 10–8 = 0.07 mSv

86

Annex III

DETERMINING THE ROUTE OF INTAKE FOR DOSE ASSESSMENT

III–1. EXAMPLE I

A worker was exposed to UF6 and UO2F2, which are classified asabsorption Type F, during his routine work. One day after conducting a specialtask, he provides 24 h samples of urine and faeces. The activities measured inthe samples were, respectively, 360 Bq/24 h of 238U and 140 Bq/24 h of 238U. Thevolume and mass of the samples of urine and faeces provided were compatiblewith the expected excretion for 24 h. Further samples of urine and faeces wereprovided for bioassay purposes two days and four days after the first sampling(days 3 and 5 after the presumed intake): the results are shown in Table III–1.

III–1.1. Route

We know the date of intake. Intake occurs mostly by ingestion, by theworker touching his or her mouth with contaminated hands, but it is necessaryto determine the route of intake in this example, in order to interpret thebioassay results in terms of intake and to calculate the dose.

Comparing the m(t) values in Appendix III for absorption Type F, 5 µmAMAD, and ingestion, f1 = 0.02, with the monitoring data, it could beconcluded that inhalation was the route of exposure. At one day afterexposure, activities in the urine were higher than in the faeces, a result thatwould not have been expected if the intake had been by ingestion. At five daysafter intake, the activities excreted in urine and in faeces were of the sameorder of magnitude, again a result that was not compatible with the ingestionroute of intake.

The worker was thus exposed to inhalation of Type F uranium. Using them(t) of Appendix III, we have the point estimates of the intake shown inTable III–2, assuming a 5 µm AMAD.

TABLE III–1. SAMPLES OF URINE AND FAECES

Days after intake Urine (Bq/24 h) Faeces (Bq/24 h)

3 12 90

5 10 12

87

The effective dose due to 238U, using the dose coefficient in Table 3, is:

E(50) = 2246 × 5.8 × 10–7

= 1.303 × 10–3 Sv, or 1.3 mSv

The activity of natural uranium is composed of 0.489 234U, 0.022 235U and0.489 238U. Thus the effective dose due to natural uranium is calculated byadding the contributions from 238U (1.3 mSv), 234U (2246 × 6.4 × 10–7 = 1.4 mSv)and 235U (((2246/0.489) × 0.022) × 6.0 × 10–7 = 0.06 mSv). Dose coefficientsfrom Table 3 were used. The total effective dose is:

E(50) = 2.8 mSv

III–2. EXAMPLE II

A worker was exposed to airborne oxides of 232Th, which is classified asType S. Air sampling in the installation, using a cascade impactor, showed thatthe AMAD was 1 µm. Routine monitoring was accomplished through thecollection of samples of faeces, and in general the results were below detectionlimits. On this occasion, however, just before going on leave and ten days aftera negative result, a worker provided a 24 h sample, and results showed anactivity of 12 Bq in the faeces. The worker recalled having had an extra load ofwork the day before he provided the sample of faeces. On the last day of his

TABLE III–2. POINT ESTIMATES OF THE INTAKE

Days after intake SampleActivity

(Bq/24 h)m(t) (from

Appendix III)

Point estimatesof the intake

(Bq)

1 Urine 360 1.8 × 10–1 2000

1 Faeces 140 5.6 × 10–2 2500

3 Urine 12 5.1 × 10–3 2353

3 Faeces 90 3.9 × 10–2 2308

5 Urine 10 4.2 × 10–3 2380

5 Faeces 12 6.2 × 10–3 1935

Average 2246

88

20 day vacation, prior to returning to work, he collected, at his home, a 24 hsample of faecal excretion. This sample was analysed and was found to bebelow the detection limit of the technique (1 mBq/24 h).

III–2.1. Route

Workers often have intakes via ingestion, from their habit of touching themouth with contaminated hands, in the work environment. The ingestedactivity causes severe interference with the monitoring results. Thus it isnecessary to evaluate the route of intake before the bioassay results are used tocalculate the dose to the worker.

III–2.1.1. Inhalation hypothesis

If the main route of intake was inhalation, one would expect, using them(t) values in Appendix III, for Type S 232Th, 1 µm AMAD, an intake of I = 12/6.1 × 10–2 = 200 Bq one day before the routine sample collection.

Using the same table of m(t), an activity of 7.6 × 10–4 × 200 = 0.15 Bq/24 hwould be expected in faeces after a 20 day vacation. This value is well above theminimum detection limit of the technique.

If the intake had occurred in the middle of the ten day interval betweensample collections, the intake would have been I = 12/8.4 × 10–3 = 1430 Bq, andthe amount excreted in faeces after the 20 day vacation would have been of theorder of 1 Bq/24 h, above the detection limit of the measuring technique.

III–2.1.2. Ingestion hypothesis

If the intake had been by ingestion one day before the vacation, themonitoring results would have corresponded to an intake of I = 12/2.8 × 10–1

= 43 Bq, using the m(t) values for ingestion (f1 = 2 × 10–3) in Appendix III. The expected amount in faeces after the vacation would have been (12/

2.8 × 10–1) × 1.5 × 10–8 = 6.4 × 10–7 Bq/24 h, a result below the detection limit ofthe technique. If the intake had occurred in the middle of the 10 day monitoringinterval, the intake would have been around 400 Bq (12/3.1 × 10–2), anunrealistic result in terms of just touching the mouth with contaminated hands.

From the above considerations it can be concluded that ingestion was themost probable exposure pathway. It is reasonable to assume that the intakeoccurred one day before the worker left on vacation, when, due to the extraload of work, the worker may have contaminated his mouth. Using thisassumption, the intake was I = 43 Bq. Using the effective dose coefficient fromTable 3, the effective dose to this worker is calculated as:

89

E(50) = 43 × 9.2 × 10–8 = 400 × 10–8

= 4 µSv

If intake by inhalation were assumed one day before the vacation, theeffective dose to the worker would have been calculated as:

E(50) = 200 × 2.3 × 10–5

= 4.6 mSv

which is a result three orders of magnitude higher than the dose calculated,assuming ingestion as the pathway of exposure.

90

Annex IV

ANALYSIS OF AN INTAKE OF MIXED ACTIVATION AND FISSION PRODUCTS FOR DOSE ASSESSMENT

IV–1. EXAMPLE

A worker who was an employee of a small specialized maintenancecompany performed maintenance work at a nuclear power plant. The workperformed was cleaning a tank using a concentrate that must be wet during theprocess. However, the work was not performed strictly according to writtenprocedures, and the man worked with dry concentrate. When he was leavingthe controlled area, surface contamination was found on his face, and subse-quently internal contamination was verified.

The first conservative estimation of internal exposure with an influence ofsurface contamination suggested that the committed effective dose could beabove the appropriate derived investigation level; whole body counts weretherefore repeatedly performed. The counts identified the corrosion products110mAg, 58Co, 60Co, 124Sb and 54Mn. In addition, analysis of excreta wasperformed.

This case is described in detail in Ref. [IV–1]. Here, in this example, onlythe interpretation of whole body measurements of 60Co is performed.

IV–2. CHARACTERISTICS OF THE INTAKE

The characteristics of the intake were:

(a) Radiation worker (male, 20 years; weight: 70 kg; height: 162 cm).(b) Intake via inhalation.(c) Date of contamination: 3 September 1998.

Table IV–1 shows the whole body count results for 60Co (date ofmeasurement and measured activity). In addition to the dates of the counts, theelapsed time in days after intake is shown. The whole body retention valuesfrom Appendix III for inhalation of a 5 µm Type S aerosol are listed incolumn 4, and the intake is calculated in column 5 by dividing the values ofcolumn 3 by the values of column 4, following the formula given in Section 3.1.

91

TABLE IV–1. WHOLE BODY COUNT RESULTS FOR 60Co

Days after intake

Measurement result (Bq)

m(t) (fromAppendix III)

Calculated intake (Bq)

4 September 1998 1 136 910 0.490 2.8 × 105

7 September 1998 4 3 588 0.098 3.7 × 104

8 September 1998 5 3 793 0.080 4.7 × 104

8 September 1998 5 3 580 0.080 4.5 × 104

9 September 1998 6 3 040 0.073 4.2 × 104

10 September 1998 7 2 978 0.069 4.3 × 104

11 September 1998 8 3 206 0.068 4.7 × 104

14 September 1998 11 2 741 0.064 4.3 × 104

15 September 1998 12 2 808 0.064 4.4 × 104

16 September 1998 13 2 440 0.063 3.9 × 104

18 September 1998 15 2 434 0.061 4.0 × 104

22 September 1998 19 2 745 0.059 4.7 × 104

23 September 1998 20 2 778 0.058 4.8 × 104

30 September 1998 27 2 415 0.055 4.4 × 104

2 October 1998 29 2 753 0.054 5.1 × 104

7 October 1998 34 2 505 0.052 4.8 × 104

9 October 1998 36 2 569 0.052 4.9 × 104

14 October 1998 41 2 564 0.050 5.1 × 104

16 October 1998 43 2 861 0.049 5.8 × 104

30 October 1998 57 2 084 0.046 4.5 × 104

4 November 1998 62 2 346 0.045 5.2 × 104

6 November 1998 64 2 083 0.044 4.7 × 104

11 November 1998 69 2 292 0.043 5.3 × 104

13 November 1998 71 2 021 0.043 4.7 × 104

20 November 1998 78 1 912 0.041 4.7 × 104

27 November 1998 85 1 993 0.040 5.0 × 104

4 December 1998 92 1 888 0.040 4.7 × 104

11 December 1998 99 1 916 0.039 4.9 × 104

18 December 1998 106 1 760 0.039 4.5 × 104

8 January 1999 127 1 767 0.037 4.8 × 104

29 January 1999 148 1 599 0.035 4.6 × 104

26 February 1999 176 1 603 0.033 4.9 × 104

92

IV–3. DOSE ASSESSMENT

From the best fit of the whole body measurements, the particle size wasderived as a 5 µm AMAD, classified as Type S. If the result of the firstmeasurement performed on the day after the intake is excluded, allmeasurement results show good agreement with the biokinetic standard model:the intakes derived from these measurement results are all within a relativelynarrow range. The arithmetic mean of these values is 46 kBq, which gives aneffective dose of:

E(50) = 46 000 × 1.7 × 10–8

= 0.78 mSv

using the dose coefficient from Table 3.It is possible to disregard the first measurement value; it indicates a

higher intake, but it is excreted very fast. This was confirmed by the faecalexcretion values obtained in this example, which are not shown here. Themodel used here is therefore not suitable for the first day after intake, since itdoes not take into account the substantial fraction of activity that is excretedvery fast; however, because of the fast excretion, the contribution to the dosefrom this fraction of activity can be neglected. The dose estimate given above istherefore in agreement with the whole body measurement results.

This is a simple example, which includes only partial data from theincident. A more complex analysis, including all other measurement results notconsidered here, may imply some modifications to the model and to the doseassessment.

26 March 1999 204 1 393 0.031 4.5 × 104

27 April 1999 236 1 084 0.030 3.6 × 104

21 May 1999 260 1 141 0.029 3.9 × 104

23 June 1999 293 935 0.027 3.5 × 104

TABLE IV–1. WHOLE BODY COUNT RESULTS FOR 60Co (cont.)

Days after intake

Measurement result (Bq)

m(t) (fromAppendix III)

Calculated intake (Bq)

93

REFERENCE TO ANNEX IV

[IV–1] FOLTÁNOVÁ, I. et al., “A case of internal contamination of a person witha mixture of radionuclides”, Radiation Protection in Central Europe (Proc.Congr. Budapest, 1999), Roland Eötvös Physical Society, Budapest (1999)496–502.

94

Annex V

DOSE ASSESSMENT FROM AN EXPOSURE OVER A PERIOD OF TIME

V–1. EVENT

An incident was discovered that had led to airborne activity of 131I in aparticular section of a workplace for a period of a few days. A worker had beenexposed the day before a weekend break and then for two days after theweekend (i.e. on Friday, Monday and Tuesday). The intakes on these days wereassumed to be of equal magnitude and were thought to have occurred overrelatively short periods of time in such a way that they could be consideredacute in nature. It was assumed that the activity was in the form of a Type Fcompound with a 5 µm AMAD. Thyroid monitoring was the assay methodselected, and measurements were carried out on the Wednesday and Thursday,which showed 480 kBq and 440 kBq, respectively.

V–2. SOLUTION

The relevant data (for ten days only) taken from the tables of m(t) givenin Appendix III are shown in Table V–1.

TABLE V–1. RELEVANT DATA

Time (days) Thyroid m(t)

1 1.20 × 10–1

2 1.20 × 10–1

3 1.10 × 10–1

4 9.90 × 10–2

5 9.00 × 10–2

6 8.20 × 10–2

7 7.40 × 10–2

8 6.80 × 10–2

9 6.20 × 10–2

10 5.60 × 10–2

95

The data given in Table V–1 can be combined to give predictions forWednesday and Thursday by introducing time offsets of two and three days forthe intakes on Monday and Tuesday, respectively, and summing horizontally(see Table V–2).

The fourth column contains the predicted values, m(t), for this multipleintake. These values can be used to estimate intakes following the methodsdescribed in Section 3.

Thyroid monitoring on Wednesday and Thursday provided the followingvalues and estimates of daily intake, I, (which are assumed to be the same foreach day):

Wednesday: 480 kBq; thus I = 480 kBq/0.330 = 1455 kBq

Thursday: 440 kBq; thus I = 440 kBq/0.312 = 1410 kBq

The estimates are consistent and a simple average of 1433 kBq can betaken to be the estimated intake for each of Friday, Monday and Tuesday.

The effective dose, using the dose coefficients from Table 3, is:

E(50) = 3 × 1433 × 103 × 1.1 × 10–8

= 47 mSv

In more complicated cases the intakes on each day may not be equal, andthe three columns of Table V–2 for Friday, Monday and Tuesday would have tobe multiplied by a suitable factor. In addition, the number of days over which

TABLE V–2. PREDICTED INTAKE VALUES

Thyroid m(t)

Intake onFriday

Intake onMonday

Intake onTuesday

Horizontal sum

Saturday 0.120 — — 0.120

Sunday 0.120 — — 0.120

Monday 0.110 — — 0.110

Tuesday 0.099 0.12 — 0.219

Wednesday 0.090 0.12 0.12 0.330

Thursday 0.082 0.11 0.12 0.312

96

exposure occurred may not be clear. Nevertheless, the principle illustrated herecan be used to calculate the dose for this kind of exposure over a number ofdays. A further example of this kind is given in Ref. [V–1].

REFERENCE TO ANNEX V

[V–1] BIRCHALL, A., HODGSON, A., MOODY, J.C., Implications of assuming arealistic intake regime for chronic exposure to airborne uranium, Radiat. Prot.Dosim. 79 (1998) 253–257.

97

Annex VI

DIRECT DOSE ASSESSMENT FOR INTAKES OF TRITIATED WATER

In Canada deuterium–uranium (CANDU) reactors and at otherworkplaces many workers are exposed chronically to low levels of HTO in theatmosphere, which results in intakes by inhalation and through the skin. Chronicor intermittent intakes of HTO at unknown times with respect to the bioassaysample is one important circumstance in which a direct dose calculation may bepreferred over other types of dose assessment (Section 3.4.2).

For both forms of intake, the HTO mixes within minutes throughout bodywater, and is excreted with the turnover of that body water. Owing to the manyroles of body water in human physiology, however, this turnover rate is highlyvariable. The 90% range of half-times in a large group of workers monitoredover several years at the Savannah River plant was 5.5 to 14.3 days [VI–1], butmuch shorter times (some three days) have been reported in hot countries[VI–2]. In this situation a biokinetic model using default parameters can begrossly unrepresentative and can lead to the underestimation or overestimationof the committed effective dose by a factor of 2 or more. For these reasons,therefore, intakes of HTO in many workplaces are best assessed by a directdose calculation [VI–3].

Tritium decays with the emission of only a weak beta particle (meanenergy 5.7 keV [VI–4]) and intakes are therefore detectable only from excretedactivity. Since urine concentrations of tritium rapidly approach those in bodywater, the dose rate to the soft tissues in which this water is distributed can beestimated from spot urine samples, taken several times per day if necessary.Using the values of E = 0.0057 MeV and m = 68.8 kg [VI–5], the specificeffective dose rate per unit activity is calculated to be:

(VI–1)

That is, the dose rate is 1.15 × 10–12 Sv per day per Bq total tritium burden, or4.8 × 10–11 Sv per day per Bq/L in the 42 L of body water. Doses over themonitoring period, usually less than two weeks, can then be estimated within50% at the 95% confidence level by linear interpolation [VI–6].

�D = ¥ ÊËÁ

ˆ¯̃

( )( )

ÊËÁ

ˆ¯̃

-1 6 10 86 4001

113.

JMeV

0.0057 MeV

68.8 kgsd BBq s

Svd Bq

◊( )

= ¥◊

ÊËÁ

ˆ¯̃

-�D 1 15 10 12.

98

From Figure VI–1, the effective dose (Sv) over the period from ti to ti+1

(days) can be estimated from the concentrations of HTO in urine, Ci (Bq/L),using the following equation:

Effective dose = 4.8 × 10–11 [(Ci+1 + Ci)/2](ti+1 – ti) (VI–2)

A simple estimate of the committed effective dose, E (Sv), resulting fromthe accumulated burden of tritium for the time period after the lastmeasurement can be derived from the final urine sample, Cn, using a defaulthalf-time to estimate continuing retention. In the absence of other evidence, ahalf-time of ten days has been recommended [VI–7]. The following equationdescribes this calculation:

(VI–3)

A small fraction (~1–3%) of an intake of HTO becomes bound to carbon intissues as a result of metabolism and is assumed to be retained with the meanhalf-time for carbon turnover, 40 days [VI–8]. Since the committed dose as aresult of this bound tritium is only about 10% of that due to the circulating HTO,and independent measurement of the bound tritium is usually not practicable, its

HTO

in u

rine

(Bq

/L)

Effe

ctiv

e d

ose

rate

(mS

v/d

ay)

t1

A B C D

C2

Monitoring period, T

C3

C4

C5

T1/2 = 10 days C1

t4t3t2 t5

Ts

Time

Te

EC

Cnn= =

--4 8 10

2 106 9 10

1110.

ln.

××

FIG. VI–1. Calculation of effective dose from chronic intakes of HTO by interpolation ofdose rates determined from the activity concentration in urine.

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effects on dose are taken into account by increasing the dose coefficients forHTO by 10% [VI–5], to 5.3 × 10–11 Sv per day per Bq/L and 7.6 × 10–10 Sv perBq/L, respectively. Although this simplified model no longer rigorously predictsthe tissue dose rate, it is accurate to within 10% during chronic exposure, andcorrectly determines the committed effective dose, since the bound tritium isexcreted well within the 50 year integration period.

As an example, Table VI–1 shows all the tritium concentrations measuredin the urine of a worker over a six week period. Table VI–1 also shows dosesreceived in each of the sampling periods, calculated in accordance with theequations above, with the dose factor adjusted as described above.

The monitoring period, T, for which total doses received are to bereported to the regulator is 1 January 2002 to 1 February 2002. The dose to bereported is calculated by apportioning the doses from those sampling periodsthat include the start and end of T. Here the total dose received in T is 1.25 mSv.The committed effective dose, calculated from the last sample using anadjusted Eq. (VI–3), is 0.76 mSv.

REFERENCES TO ANNEX VI

[VI–1] BUTLER, H.L., LEROY, J.H., Observation of the biological half-life of tritium,Health Phys. 11 (1965) 283–285.

[VI–2] RUDRAN, K., Significance of in-vivo organic binding of tritium followingintake of tritiated water, Radiat. Prot. Dosim. 25 (1988) 5–13.

[VI–3] HEALTH AND WELFARE CANADA, Bioassay Guideline 2, Guidelines forTritium Bioassay 83-EHD-87, Canadian Department of Health and Welfare,Ottawa (1983).

TABLE VI–1. TRITIUM CONCENTRATIONS MEASURED IN URINEOF WORKER

Tritium concentration in urine (MBq/L)

Effective dose insampling period (mSv)

28 December 2001 0.40 —

5 January 2002 1.20 0.34

14 January 2002 0.70 0.45

25 January 2002 0.50 0.35

8 February 2002 1.00 0.56

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[VI–4] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, Radionuclide Transformations: Energy and Intensity of Emissions,Publication 38, Pergamon Press, Oxford and New York (1983).

[VI–5] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, Age-dependent Doses to Members of the Public from Intake ofRadionuclides: Part 2, Ingestion Dose Coefficients, Publication 67, PergamonPress, Oxford and New York (1993).

[VI–6] HEALTH PHYSICS SOCIETY, Internal Dosimetry Programmes for TritiumExposure — Minimum Requirements, American National Standard HPSN13.14, Health Physics Society, McLean, VA (1994).

[VI–7] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, Reference Man: Anatomical, Physiological and MetabolicCharacteristics, Publication 23, Pergamon Press, Oxford and New York (1975).

[VI–8] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, Age-dependent Doses to Members of the Public from Intake ofRadionuclides: Part 1, Publication 56, Pergamon Press, Oxford and New York(1989).

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Annex VII

ANALYSIS OF A SINGLE INTAKE OF 238,239,240Pu AND 241Am FOR DOSE ASSESSMENT

VII–1. BACKGROUND

This annex is based on Case 8 of the IAEA Co-ordinated ResearchProject on Intercomparison and Biokinetic Model Validation of RadionuclideIntake Assessment [VII–1]. In this example the use of the tables in this reportis demonstrated, as well as the limitations of the results obtained.

On 24 May 1983 at 16:15 there was an explosion in a glovebox of a radio-chemical laboratory for the development of advanced nuclear fuels in a nuclearresearch centre. The pressure of the explosion opened the sluice of the box anddestroyed the box gloves; two workers received contamination on their faces,hair and clothes.

The activity composition of the inhaled substance was 9% 238Pu, 55%239Pu, 26% 240Pu and 10% 241Am. The diameter of the plutonium containingparticles is assumed to have been between 3 and 40 µm. The chemical form wasa hydroxide gel in washing water containing 10% ammonium nitrate and about3.5% hexamethylentetramine.

Measurements of plutonium and americium activity in body regions andin excreta for both workers were started immediately and continued over manyyears. In the example given in this annex, only some measurement resultsobtained for one of the individuals (a 26 year old male who weighed 80 kg) areused.

Measurements were taken of the 241Am lung burden beginning on thedate of the accident. The results are illustrated in Table VII–1 (with anuncertainty of 25% for each value).

Additionally, as shown in Table VII–2, two sets of detailed measurementsto determine the activity from 241Am in the lymph nodes, lungs, bone and liverwere performed in two different institutions, on 3 August 1993 in laboratory Aand on 15 November 1993 in laboratory B.

The uncertainties of the measurement results were between 12% (bone)and 16% (liver) in laboratory A and between 12% (bone) and 33% (liver) inlaboratory B.

There were several excretion measurements for 239Pu + 240Pu, as well asfor 241Am + 238Pu (Tables VII–3 and VII–4).

Additionally, two measurements to determine the urinary excretion of241Am were conducted on 25 April 1990 and 25 May 1991; these showed valuesof 4.3 and 2.3 mBq/day, respectively.

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Four further measurements of faecal excretion for 241Am were conductedon 3 May 1988, 27 August 1988, 24 April 1990 and 25 May 1991. These yielded0.018, 0.025, 0.012 and 0.0056 mBq/day, respectively.

It was necessary to determine the intake of 241Am and of the plutoniumisotopes. In addition, the committed effective dose and the committed dose tothe most highly exposed organ, the bone surfaces, needed to be calculated.

TABLE VII–1. AMERICIUM-241 ACTIVITY IN THE LUNGS

Activity in the lungs (Bq)

24 May 1983 390

25 May 1983 310

27 May 1983 230

8 June 1983 230

27 June 1983 230

1 July 1983 260

7 July 1983 230

31 October 1983 220

4 November 1983 230

15 May 1984 220

5 May 1986 240

27 May 1991 180

TABLE VII–2. AMERICIUM-241 ACTIVITY IN OTHER ORGANS

Organ activity (Bq)

Laboratory A Laboratory B

Lymph nodes 26 72

Lungs 120 120

Bone 69 65

Liver 57 24

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VII–2. DOSE CALCULATION

It is evident from the lung activity data that there was a remarkabledecrease of lung activity within the first days, followed by a plateau over aperiod of three years and only a slight decrease later on. This shows that thebehaviour of americium in the lungs is much closer to that of absorption Type Sthan to that of absorption Type M. The problem in using the tables in thisreport is that the ICRP in Ref. [VII–2] (and therefore also in Ref. [VII–3]) onlyconsiders Type M (moderate absorption), while the actual data show a muchslower absorption. Such a slower absorption is, however, also known fromassessments of other cases [VII–4].

In this example we have the possibility of using the lung retention data of239Pu, for which Type S (slow absorption) is given in the tables. Both 241Am and239Pu have a very long half-life compared with the time period observed here,and therefore the lung retention times are quite similar. In the tables of intakeretention fractions given in Appendix III, values of 1 and 5 µm are given for the

TABLE VII–3. URINARY EXCRETION RATES FOR PLUTONIUMAND AMERICIUM

Urinary excretion rate (mBq/day)

239Pu + 240Pu 241Am + 238Pu

25 May 1983 11 110

26 May 1983 41 100

7 June 1983 4.7 16

14 June 1983 3.7 11

24 June 1983 3.7 5.6

30 June 1983 5.6 5.6

6 July 1983 3.7 5.2

21 November 1983 3.7 4.6

26 May 1984 3.5 4.0

20 January 1985 2.9 3.4

3 May 1986 3.7 2.7

27 August 1988 5.9 4.7

11 February 1989 6.2 3.8

28 January 1994 3.4 2.6

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AMADs. Since the AMAD in this example is given to be between 3 and 40 µm,the higher value seems to be the more appropriate.

In Table VII–5 the measurement values of the lung activity (in Bq) atspecified times after intake (in days) are given with the appropriate lungretention values from the data given in this report. From this, for eachmeasurement the intake (in Bq) is calculated. The first measurement valueshortly after the intake is not used here because appropriate retention valuesshortly after intake are not given and because the exact time of measurement isnot given.

An intake of about 4–5 kBq of 241Am can be assumed from the first sixvalues. The exact intake value depends on the assumed AMAD, whichinfluences the activity fraction deposited in the lungs, especially in the deepparts of the lungs, which have a long retention period. It can be seen from thesecalculations that even the assumption of a Type S material underestimates theretention in the lungs at long times after intake, giving very high intake valuesfrom these measurement results.

For the evaluation of the excretion measurements for 241Am we have theproblem that the ICRP model for americium, and therefore the data given inthis report, only considers a Type M behaviour for lung absorption, while weare considering here a Type S behaviour. The biokinetic behaviour of

TABLE VII–4. FAECAL EXCRETION RATES FOR PLUTONIUMAND AMERICIUM

Faecal excretion rate (Bq/day)

239Pu + 240Pu 241Am + 238Pu

25 May 1983 5200 1500

26 May 1983 3000 740

27 May 1983 440 74

6 June 1983 0.67 0.16

14 June 1983 0.72 0.15

23 June 1983 0.67 0.12

30 June 1983 0.25 0.078

7 July 1983 0.21 0.059

21 November 1983 0.42 0.094

27 May 1983 0.26 0.059

20 January 1985 0.26 0.075

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plutonium is similar, but not identical, to that of americium, and therefore theType S values for plutonium cannot be used for dose assessment.

Tables VII–6 and VII–7 show the urinary and faecal excretion data for239Pu + 240Pu, respectively. Both isotopes have a long half-life and can beconsidered together. As for the americium estimation used in this example, aType S lung absorption and an AMAD of 5 µm are assumed.

The model urinary excretion function with an assumed intake of 25 kBqfor both plutonium isotopes shows good agreement with all urine measure-ments. The faecal excretion function shows good agreement with themeasurement values only for the early measurements and for the measure-ments made half a year after intake and thereafter if an intake of 10 kBq isassumed for the two plutonium isotopes. However, in this example the urinarymeasurement results seem to be more reliable for dose assessment, since thefaecal excretion is very strongly influenced by the lung retention and themucociliary transport into the alimentary tract. The intake of 25 kBq for bothplutonium isotopes also agrees rather well with the assessed 241Am intake of4–5 kBq, keeping in mind that the total activity of the plutonium isotopes isabout eight times the activity of 241Am.

Evaluation of the organ measurements on 241Am activity is not possiblewith the material given in this report, since the organ retention functions areonly given for Type M material, not for Type S material. For an evaluation it

TABLE VII–5. ESTIMATED INTAKE VALUES BASED ONMEASUREMENTS AT SPECIFIED TIMES AFTER INTAKE

Days after intakeLung activity

(Bq)m(t)

(from Appendix III)Intake (Bq)

1 310 6.4 × 10–2 4.8 × 103

3 230 6.2 × 10–2 3.7 × 103

15 230 5.5 × 10–2 4.2 × 103

34 230 4.9 × 10–2 4.7 × 103

38 260 4.8 × 10–2 5.4 × 103

44 230 4.5 × 10–2 5.1 × 103

160 220 3.3 × 10–2 6.7 × 103

164 230 3.3 × 10–2 7.0 × 103

357 220 2.7 × 10–2 8.1 × 103

1077 240 1.4 × 10–2 1.7 × 104

2925 180 8.5 × 10–3 2.1 × 104

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would be necessary to calculate these values with the assumed modelparameters. However, further modifications would be necessary because, ashas been mentioned, the model lung retention function does not agree with themeasurement values for extended periods after intake. Therefore, withoutfurther modifications to the model, we would obtain incorrect intake estimates.

The doses caused by plutonium isotopes could, if the biokinetic modelsare not further modified, be calculated by using the effective dose coefficientsgiven in Table 3; for example, with an assumed intake of 25 kBq of 239Pu +240Pu, we would obtain a committed effective dose of 210 mSv and a committedequivalent dose to bone surfaces of 2.3 Sv (the bone surface dose coefficient forintake of a 5 µm AMAD Type S aerosol for both 239Pu and 240Pu is 9.1 × 10–5

[VII–5]). However, it must be kept in mind that, for example, the lung dose isprobably underestimated, since the lung measurement data indicate a muchlonger retention than due to the standard Type S assumptions used here, whichalso bear on the effective dose. To assess a dose resulting from an intake of241Am the dose coefficients given in this report cannot be used, since they are

TABLE VII–6. ESTIMATED INTAKE VALUES BASED ONEVALUATION OF URINARY EXCRETION DATA

Days after intake

Urinary excretion (Bq/day)

Excretion rate per unit intake (m(t))

Intake (kBq)

25 May 1983 1 1.10 × 10–2 2.3 × 10–6 4.7

26 May 1983 2 4.10 × 10–2 1.4 × 10–6 30

7 June 1983 14 4.70 × 10–3 2.1 × 10–7 25

14 June 1983 21 3.70 × 10–3 1.8 × 10–7 21

24 June 1983 31 3.70 × 10–3 1.7 × 10–7 22

30 June 1983 37 5.60 × 10–3 1.7 × 10–7 33

6 July 1983 43 3.70 × 10–3 1.7 × 10–7 22

21 November 1983 181 3.70 × 10–3 1.6 × 10–7 23

26 May 1984 368 3.50 × 10–3 1.7 × 10–7 21

20 January 1985 607 2.90 × 10–3 1.8 × 10–7 17

3 May 1986 1075 3.70 × 10–3 1.8 × 10–7 21

27 August 1988 1922 5.90 × 10–3 1.6 × 10–7 37

11 February 1989 2090 6.20 × 10–3 1.6 × 10–7 40

28 January 1994 3902 3.40 × 10–3 1.2 × 10–7 29

107

not given for a Type S lung absorption. It would therefore be necessary tocalculate them, which would require a major effort.

This annex attempts to give solutions to a real incorporation case usingthe material given in this report. As for many other cases, it can be seen herethat the standard models do not apply. However, with some careful assump-tions, some dose assessments can be achieved. Either way, the kinds of errorthat are introduced by the use of models that are not very appropriate for thecase given must be considered.

For a very rigorous dose assessment, powerful tools and expertexperience are needed to adapt the models to the particular situation and tocalculate intake and dose values using these individually modified models.

REFERENCES TO ANNEX VII

[VII–1] INTERNATIONAL ATOMIC ENERGY AGENCY, Intercomparison andBiokinetic Model Validation of Radionuclide Intake Assessment — Report ofa Co-ordinated Research Project 1996–1998, IAEA-TECDOC-1071, IAEA,Vienna (1999).

TABLE VII–7. ESTIMATED INTAKE VALUES BASED ONEVALUATION OF FAECAL EXCRETION DATA

Days afterintake

Faecal excretion(Bq/day)

Excretion rate per unit intake (m(t))

Intake(kBq)

25 May 1983 1 5.20 × 103 1.1 × 10–1 46

26 May 1983 2 3.00 × 103 1.6 × 10–1 18

27 May 1983 3 4.40 × 102 8.4 × 10–2 5.2

6 June 1983 13 6.70 × 10–1 5.9 × 10–2 1.3

14 June 1983 21 7.20 × 10–1 4.3 × 10–4 1.7

23 June 1983 30 6.70 × 10–1 3.5 × 10–4 1.9

30 June 1983 37 2.50 × 10–1 3.0 × 10–4 0.830

7 July 1983 44 2.10 × 10–1 2.6 × 10–4 0.810

21 November 1983 181 4.20 × 10–1 4.5 × 10–5 11

27 May 1984 369 2.60 × 10–1 2.2 × 10–5 12

20 January 1985 607 2.60 × 10–1 1.7 × 10–5 15

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[VII–2] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, Dose Coefficients for Intakes of Radionuclides by Workers, Publication68, Pergamon Press, Oxford and New York (1994).

[VII–3] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, Individual Monitoring for Internal Exposure of Workers, Publication 78,Pergamon Press, Oxford and New York (1997).

[VII–4] NOßKE, D., RÜHM, W., KARCHER, K., Individual dose assessment ofworkers, Radiat. Prot. Dosim. 79 (1998) 83–86.

[VII–5] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTEC-TION, The ICRP Database of Dose Coefficients: Workers and Members of thePublic, Elsevier, Oxford (1999) (CD-ROM).

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Annex VIII

CHOOSING THE APPROPRIATE MONITORING PERIODFOR DOSE ASSESSMENT

VIII–1. BACKGROUND

A radiation protection officer of an installation was assigned the duty ofestablishing the schedule and best bioassay monitoring technique for workersexposed to elemental 131I, classified as vapour. The monitoring frequency hadto be chosen in such a way as to detect an intake occurring at the beginning ofthe monitoring period and corresponding to one tenth of the annual effectivedose limit of 20 mSv. There was a further condition that the uncertainty in thecalculated intake, because of the unknown time of intake, was to be less than afactor of two. The MDA for the in vivo measurement facility in the installationwas 40 Bq for 15 min of thyroid monitoring. The MDA for iodine in urine wasfound at the in vitro laboratory to be 1 Bq/L.

VIII–2. CALCULATION

VIII–2.1. To comply with the first condition

The effective dose coefficient for 131I was 2 × 10–8 Sv/Bq (Table 3). Theintake to be detected by the chosen monitoring technique corresponded to:

I = [(1/10 × 20 × 10–3)/2 × 10–8] = 1 × 105 Bq

For urine monitoring:

MDA = 1Bq/L

m(t) = 1/(1 × 105) = 1 × 10–5

On the basis of the m(t) table in Appendix III, a maximum monitoringinterval of 50 days was advisable.

For in vivo monitoring:

MDA = 40 Bq

m(t) = 40/(1 × 105) = 4 × 10–4

110

On the basis of the m(t) table in Appendix III, a maximum monitoringinterval of 70 days was advisable.

VIII–2.2. To comply with the second condition

On the basis of the m(t) table in Appendix III for urine, urine monitoringis not appropriate for 131I vapour. The difference in intake assessment fromday 1 to day 2 after exposure is more than a factor of 2.

On the basis of the m(t) table in Appendix III for the thyroid, for 131Ivapour a maximum monitoring interval of ten days was advisable.

VIII–3. CONCLUSION

In vivo monitoring with a ten day interval was recommended for theinstallation.

111

GLOSSARY

acute intake. An intake occurring within a time period short enough that it canbe treated as instantaneous for the purposes of assessing the resultingcommitted dose.

bioassay. Any procedure used to determine the nature, activity, location orretention of radionuclides in the body by direct (in vivo) measurement orby in vitro analysis of material excreted or otherwise removed from thebody.

biokinetic model. A mathematical model describing the intake, uptake andretention of a radionuclide in various organs or tissues of the body andthe subsequent excretion from the body by various pathways.

biological half-life. The time taken for the quantity of a material in a specifiedtissue, organ or region of the body (or any other specified biota) to halveas a result of biological processes.

chronic intake. An intake over an extended period of time, such that it cannotbe treated as a single instantaneous intake for the purposes of assessingthe resulting committed dose.

derived air concentration (DAC). A derived limit on the activity concentrationin air of a specified radionuclide, calculated such that Reference Man,breathing air with a constant contamination at the DAC while performinglight physical activity for a working year, would receive the annual limiton intake for the radionuclide in question.

fractional absorption in the gastrointestinal tract (f1). The f1 value is thefraction of an element directly absorbed from the gut to body fluids.

intake. The act or process of taking radionuclides into the body by inhalation oringestion or through the skin. Or the activity of a radionuclide taken intothe body in a given time period or as a result of a given event.

minimum detectable activity (MDA). The activity which, if present in a sample,produces a counting rate that will be detected (i.e. considered to be abovebackground) with a certain level of confidence.

112

The “certain level of confidence” is normally set at 95%, i.e. a samplecontaining exactly the MDA will, as a result of random fluctuations, betaken to be free of activity 5% of the time.The MDA is sometimes referred to as the detection limit or lower limit ofdetection. The counting rate from a sample containing the MDA istermed the determination level.

minimum significant activity (MSA). The activity which, if present in a sample,produces a counting rate that can be reliably distinguished frombackground with a certain level of confidence.A sample containing exactly the MSA will, as a result of random fluctua-tions, be taken to be free of activity 50% of the time, whereas a truebackground sample will be taken to be free of activity 95% of the time.The MSA is sometimes referred to as the decision limit. The counting ratefrom a sample containing the MSA is termed the critical level.

radioactive half-life. For a radionuclide, the time required for the activity todecrease by a radioactive decay process, by half.

transfer compartment. The compartment introduced for mathematicalconvenience into most of the biokinetic models used in ICRP and IAEApublications to account for the translocation of the radioactive materialthrough the body fluids from where they are deposited in tissues.

uptake. The processes by which radionuclides enter the body fluids from therespiratory tract, gastrointestinal tract or through the skin, or the fractionof an intake that enters the body fluids by these processes.

Human Respiratory Tract Model (HRTM):

activity median aerodynamic diameter (AMAD). The value ofaerodynamic diameter1 such that 50% of the airborne activity in aspecified aerosol is associated with particles smaller than the AMAD and50% of the activity is associated with particles larger than the AMAD.

1 The aerodynamic diameter of an airborne particle is the diameter that a sphereof unit density would need to have in order to have the same terminal velocity whensettling in air as the particle of interest.

113

• Used in internal dosimetry for simplification, as a single ‘average’ valueof aerodynamic diameter representative of the aerosol as a whole.

• The AMAD is used for particle sizes for which deposition dependsprincipally on inertial impaction and sedimentation: typically thosegreater than about 0.5 µm. For smaller particles, deposition typicallydepends primarily on diffusion, and the activity median thermo-dynamic diameter (AMTD) — defined in an analogous way to theAMAD, but with reference to the thermodynamic diameter2 of theparticles — is used.

activity median thermodynamic diameter (AMTD). See activity medianaerodynamic diameter (AMAD).

alveolar–interstitial (AI) region. The respiratory bronchioles, alveolar ductsand sacs with their alveoli, and the interstitial connective tissue.

bronchial (BB) region. The trachea and bronchi.

bronchiolar (bb) region. The bronchioles and terminal bronchioles.

clearance.3 The removal of material from the respiratory tract by particletransport and by uptake.

deposition. The initial processes determining how much of a material in inhaledair remains in the respiratory tract after exhalation. Deposition ofmaterial may occur during both inhalation and exhalation.

extrathoracic (ET) airways. The anterior part of the nose (ET1) and theposterior part of the nasal passages, mouth, pharynx and larynx (ET2).

particle transport. Processes that clear material from the respiratory tract tothe gastrointestinal tract and to the lymph nodes and move material fromone part of the respiratory tract to another.

thoracic (TH) airways. The bronchial (BB), bronchiolar (bb) and alveolar-interstitial (AI) regions.

2 The thermodynamic diameter of an airborne particle is the diameter that asphere of unit density would need to have in order to have the same diffusion coefficientin air as the particle of interest.

3 In order to avoid ambiguity the definition given in Ref. [15] is used here.

114

types of materials. Categories of materials in the lung according to their rates ofabsorption from the respiratory tract to body fluids:

Type F: deposited materials that are readily absorbed into body fluidsfrom the respiratory tract. (Fast rate of absorption.)Type M: deposited materials that have intermediate rates of absorptioninto body fluids from the respiratory tract. (Moderate rate of absorption.)Type S: deposited materials that are relatively insoluble in the respiratorytract. (Slow rate of absorption.)Type V: deposited materials that are assumed, for dosimetric purposes, tobe instantaneously absorbed into body fluids from the respiratory tract —applied only to certain gases and vapours. (Very rapid absorption.)

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CONTRIBUTORS TO DRAFTING AND REVIEW

Andrasi, A. KFKI Atomic Energy Research Institute, Hungary

Bailey, M. National Radiological Protection Board, United Kingdom

Bertelli, L. Argonne National Laboratory, United States of America

Castellani, C.-M. ENEA, Italy

Cruz Suárez, R. International Atomic Energy Agency

Guilmette, R.A. Los Alamos National Laboratory, United States of America

Gustafsson, M. International Atomic Energy Agency

Hurtgen, C. SCK•CEN, Belgium

Ishigure, N. National Institute of Radiological Sciences, Japan

Kramer, G.H. National Calibration Reference Centre for In Vivo Monitoring, Canada

LeGuen, B. EDF-GDF, France

Lipsztein, J. Instituto de Radioproteção e Dosimetria, IRD/CNEN, Brazil

Malátová, I. National Radiation Protection Institute, Czech Republic

Noßke, D. Bundesamt für Strahlenschutz, Germany

Perrin, M.-L. IPSN/DPHD/SEGR/SAER, France

Phipps, A. National Radiological Protection Board, United Kingdom

Stather, J.W. National Radiological Protection Board, United Kingdom

Thériault, B. Canadian Nuclear Safety Commission, Canada

Toohey, R.E. Oak Ridge Institute for Science and Education, United States of America

Whillans, D. Ontario Power Generation, Canada

Xia, Y. China Institute of Atomic Energy, China

Consultants Meetings

Vienna, Austria: 8–12 December 1997, 11–15 May 1998, 7–10 December 1998, 17–18 February 2000, 18–22 June 2001